Development of high-frame-rate neutron radiography and quantitative measurement method for multiphase flow research

In relation to the establishment of thermal neutron radiography as a measurement method with high accuracy and reliability, this paper reviewed the present status on the development of high-frame-rate neutron radiography with a steady thermal neutron beam and its application to multiphase flow researches. This review included also the present progress on the quantification of neutron radiographic image at Kyoto University, i.e. (1) quantitative method to measure void fraction of two-phase flow with thermal neutron radiography (Σ-scaling method), (2) influence of scattered neutrons on void fraction measured by neutron radiography, (3) measurement error of neutrons in a low neutron flux field, (4) error in void fraction measurement due to low gray level, and (5) measurement error due to low imaging speed Moreover, a new experimental approach on a total macroscopic cross section for thermal neutrons measurement by neutron radiography was presented. This paper revealed neutron radiography to be a promising visualization and measurement method in thermal hydraulic research.     

Analysis of Heavy-Water-Moderated,Cluster-Type Fuel Lattices by Cluster Physics Code ‘MESSIAH’

This paper presents a comprehensive analysis performed by a new cluster analysis code ‘MESSIAH’ on reactor physics constants measured in the critical facility for a pressure-tube-type, heavy-water-moderated reactor. The MESSIAH code system utilizes the method of the collision probability to solve the neutron transport equation. The effective space dependent cross sections are calculated in the thermal and resonance energy range before the eigenvalue calculation for the whole energy range. With use of these cross sections, the multi-group, space dependent transport equation is solved, and the flux distribution, spectrum and k _eff are obtained to the input bucking. In the above three steps the method of the collision probability is used consistently and extensively. The treatment of leakage neutrons from lattices in MESSIAH is also confirmed by an independent method using a Monte Carlo calculation. The calculated reactor physics constants, especially the micro-parameters and the activation traverse of Dy, agreed fairly well with the experiment. The diffusion calculation with use of the group constants calculated by MESSIAH predicts the reactivity of 0% void core excellently (<0.12%). However, for a 100% void core, the calculated reactivity was slightly lower than the experiment (~0.74%), which was attributed to over prediction of the diffusion constants.     

Determination of Large Negative Reactivity by Integral Versions of Various Experimental Methods

The large negative reactivity is measured in Semi-Homogeneous Experimental facility (SHE). Experimental methods are Sjöstrand's pulsed neutron, source multiplication and rod drop methods beside revised King-Simmons' pulsed neutron methods. Neutron detectors are placed at various points in the core region for multi-points measurement.

Usual one-point reactor model analysis resulted in the reactivity values, strongly dependent on the detector position with the increase of subcriticality. In addition, disagreements between the used experimental methods are also pointed out.

In order to overcome these difficulties due to the spatial higher harmonics and the kinetic distortion in the neutron flux distribution, an integral version analysis is applied, in which use is made of multi-points reactor model. In the analysis, space integration of the neutron counts obtained throughout the core region is made with weights of the adjoint function of fast neutrons, calculated using the two- or three-dimensional diffusion code. The negative reactivity values determined by the integral version analysis agreed well with each other within the uncertainty of ～5% in the reactivity range down to ～50 dollars.

It is concluded that all the experimental methods are adequate for precise determination of the large negative reactivity of reactor provided that the integral version analysis is utilized or that correction is made for the change of the neutron generation time using precise calculation.     

Physics Studies on BWR Void Simulated Cores, (I)

In an attempt to obtain more information concerning steam void effects on the physical behavior of BWR cores, voids ranging from zero to 24% volume fractions were statically simulated by inserting polyethylene tubes between fuel rods of simulated BWR core, and measurements were made of local flux distribution, neutron spectrum, reactivity, reactivity coefficient of void and so forth. The results were compared with calculations. Improved methods of calculation were developed, which assured better agreement with measured results.

While this paper only deals with local distributions of the thermal neutron flux and spectrum within the assembly cells, it has been shown that spectrum variation near water gaps could be well described by an overlapping model which provided an improvement on the analysis. An additional improvement was obtained by the incorporation of an approximate transport correction.     

Effects of Spatial Harmonics on Power Distribution in Accelerator Driven Subcritical Cores

A method for solving the time-energy-dependent diffusion equation was devised to appreciate the effect of spatial harmonics on the power distribution in a subcritical graphite-moderated core with a pulsed fast neutron source. An instantaneous power peaking factor (IPF) was calculated using the thermal neutron flux thus obtained, to characterize the power distribution in the core. In case of a 50-cycles-per-second injections of fast neutron pulses of 3 ms pulse width, it was indicated that the value of IPF increased by 4.1 times with the decrease in the multiplication factor from 0.9951 to 0.9762, accompanying fuel burnup. To appreciate the dependence of the core power distribution on the spectrum hardening of the thermal neutron flux caused by injections of pulsed fast neutrons, comparison was made between the value of IPF and that derived from the time-dependent-one-group diffusion theory where the spectrum hardening of the thermal neutron flux was ignored. It was indicated that the disregard of the spectrum hardening of the thermal neutron flux resulted in the low estimate of IPF that changed from ?1% to ?14% with the decrease in the pulse width of pulsed fast neutrons from 3 ms to 0.15 ms.     

Thermal Neutron Spectra in Light Water Poisoned with Cadmium and/or Indium

The thermal neutron spectra in light water of slab geometry poisoned with Cd and/or In were measured by the time of flight method, using a 20-MeV electron linear accelerator. The thermal spectra were simulated to those in the Pu built-up core of a commercial light water reactor corresponding to a fuel burnup of about 15,000 MWD/T. The results of measurements were compared with calculations based on the S ₄ method using the Haywood scattering law. Fairly good agreement was obtained between the calculated and measured results except in a limited range of energy above the 0.176 eV resonance of Cd. It is concluded that the P ₁ components of the source neutrons as well as the neutron scattering kernel play a significant role in the calculation of the thermal neutron spectra with large flux gradients, and that the scattering kernel of light water based on the Haywood model will be accurate enough to evaluate the infinite multiplication constant k_∞ of light water reactor cores with high fuel burnup within an error of about — 0.17%, as estimated from the uncertainty in the spectrum calculation in the region above the Cd resonance. It is also emphasised from the two- dimensional S ₄ calculations that the effect of reentrant hole perturbation should be evaluated quantitatively in the interpretation of the measured angular neutron spectra produced within finite media.     

Beam Hole Configuration and Analysis of Medical Reactor for Boron Neutron Capture Therapy

This report presents an investigation of beam holes to be provided in a medical reactor for Boron Neutron Capture Therapy. The principal requirement for the beam holes is to deliver the therapeutic doses of thermal and epithermal neutrons in a modest time (30 to 60min) with minimal fast neutron and γ-contaminants. Characteristics of the beam holes have been evaluated by 2-dim. n-γ coupling S _N transport calculations. Reexaminations and revisions of the beam hole design have brought improvements of the characteristics, especially an increase of the thermal neutron flux at the horizontal thermal neutron beam port and a decrease of the fast neutron flux at the vertical epithermal neutron beam port. The design objectives for the beam holes set up in this study may be achievable even if the thermal power of the reactor is reduced from 2 to 1MW.     

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract

The behavior of neutrons in a highly heterogeneous unit cell consisting of D₂O moderator, H₂O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different ²³⁵U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D₂O level, as well as various lattice parameters

A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H₂O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H₂O coolant.

The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D₂O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p ²⁸—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions.     

Development of Zircaloy Clad Hafnium Rods for BWR Long Life Neutron Absorbers

Control rod columns of the High Temperature Engineering Test Reactor (HTTR) core have each three cylindrical void holes for insertion of the control rods. The reactivity worths of the simulated void holes were measured in the Very High Temperature Reactor Critical Assembly (VHTRC) by the pulsed neutron method with the purpose of evaluating the neutron streaming effect on the reactivity worths of the void holes. The reactivity worths were determined using the revised King-Simmons' formula. The measured results are compared with the calculated ones based on the nuclear data of ENDFIB-IV, using the Benoist's anisotropic neutron diffusion coefficients. It is concluded that the reactivity worths are enlarged up to about twice as large as that obtained using the conventional isotropic neutron diffusion coefficients, and it is also found that the neutron streaming effect that is defined as the relative fraction components exceeding the reactivity worths obtained by the conventional isotropic neutron diffusion coefficients are likely overestimated about 11 and 32% with the void holes being in the core and reflector regions, respectively. Then, the reactivity decrease due to the neutron streaming effect is able to be evaluated around 1%Δk for the initial critical core of the HTTR.     

Measurement of Reactivity Effect and Thermal Neutron Flux in Non-uniformly Distributed Fuel Assemblies

A reactivity effect due to spatial variation of nuclear fuel concentration is an important problem for nuclear criticality safety in a reprocessing plant. As a method theoretically estimating this reactivity effect, the Goertzel's necessary condition is well known. To investigate this theoretical prediction, we have performed systematically a series of measurements of reactivity effect due to nonuniformity in the fuel distribution using a solid moderated core of the Kyoto University Critical Assembly. Thermal neutron flux distributions have been also measured using the activation method with gold wire. A nonuniform assembly was found to have an excess reactivity of 0.2%Δk/k when it had the same uranium mass as the uniform assembly was exactly critical. The fuel concentration of this nonuniform assembly was 40% higher in the center region than the uniform one. Moreover, the spatial distribution of reaction rate of gold wire due to thermal neutrons was flatter in this nonuniform assembly than in the uniform one, as expected by the Goertzel's condition.     

Design of a permanent epithermal neutron irradiation site in the Syrian MNSR

A permanent epithermal neutron irradiation site was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using cadmium as a thermal neutron shielding material. This site was designed by Cd-shielding the internal surface of the outer aluminum tube of the FOIS (First Outer Irradiation Site) in the MNSR. The MCNP-4C calculations showed that, to have a permanent epithermal neutron irradiation site for the ENAA using the cadmium, it is necessary to add the top beryllium shims of the reactor to compensate for the reactivity losses due to the neutrons absorption in the cylindrical cadmium shell. The activation detectors were used to measure the thermal and epithermal neutron fluxes in the FOIS. Distribution of the thermal neutron flux along the vertical direction of the outer irradiation capsule used in the FOIS has been found using MCNP-4C code, and experimentally by irradiating five copper wires. Good agreements were obtained between the calculated and the measured results.     

Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

This work is concerned with the study of the distribution and attenuation of doses of thermal neutrons emitted directly from the core of ²³⁵U research reactor in ordinary concrete shields. In practice it is not possible to identify the reactor thermal neutrons in the emitted continous neutron spectrum, therefore, measurements were carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li₂B₄O₇:Mn thermoluminescent dosimeters.The data obtained were analysed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves were constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc collimated beam and infinite plane monodirectional sources.     

A stochastically perturbed nonlinear point reactor model

We present an exact analysis of the fluctuations of neutron density in a stochastically perturbed nonlinear point reactor model in the absence of delayed neutrons. The reactivity and feedback coefficients are assumed to have white noise Gaussian component. The time development of probability distribution is obtained from the Fokker-Planck equation. It is found that for small density of neutron population the stationary probability distribution is more sensitive to the random change in the reactivity than in the feedback coefficient, whereas for large neutron density the probability distribution is more sensitive to the random change in the feedback coefficient than in the reactivity.     

Parametric analysis of the neutron transmutation doping (NTD) facility at the Egyptian Second Research Reactor (ETRR-2)

One of the most important facilities of the Egyptian Second Research Reactor (ETRR-2) is the silicon Neutron Transmutation Doping (NTD) Facility. This work is done to optimize the performance of the ETRR-2 NTD facility based on the standards and to develop a calculation line that can be used during its commercial operating phase. Calculation line methodology was developed using WIMS-D4 and CITVAP codes as well as nuclear data file libraries to perform core calculations needed to predict neutronic parameters at the two NTD locations. Axial flux profile was calculated and results were in good agreement with the experimental flux mapping. Based on the calculated variation of the thermal neutron flux axial distribution, it was proposed to displace the silicon ingot axially from its initially proposed position to minimize flux variation. Impact of control rod configuration was analyzed to maximize the neutron flux at the NTD two locations and the best control rod configuration is recommended. Thermal to fast neutron flux was also calculated and concluded that radiation damage caused by fast neutrons can safely be neglected. Flux depression across the ingot was estimated using the ANISN code. Finally, the SIMULINK module of the MATLAB code was used to evaluate the ingot activity due to different radioactive nuclides. The activity was found to reach its exempted limit after 2.79 days after its discharge from the irradiation channel. It was also concluded that the worse ETRR-2 scram will not affect the obtained Si doping and its resistivity.     

Void induced reactivity in a mixed MTR core

Experiments in research reactors often pose the necessity of void insertion in several core positions. Also, sub-cooled boiling may introduce fluctuating quantities of bubbles. The reactivity change and its effect on the safety are important matters that must be assessed. In the present work, the reactivity induced by void insertion in various locations of the pool type, mixed (high and low enrichment uranium fuel) core of the Greek Research Reactor, was measured. The induced reactivity due to the void introduction was determined by measuring the reactor period. The measured reactivity was found to be strongly space-dependent. Results obtained with the diffusion code CITATION were found in good agreement with the experiment. Conclusions on the best utilization of diffusion codes when calculating local void effects on reactivity, in regions of steep flux gradients, were drawn. The sensitivity of the local void reactivity to the core position, as well as to the fuel enrichment and burn-up was examined. An analytical point reactor model with two neutron energy groups was developed, to investigate qualitatively the effects of parameters related to void reactivity. The analytical results were found in good qualitative agreement with measurements and numerical simulations.     

Effects of Non-uniform Moderator Temperature on Core Power Distribution in Accelerator Driven Reactors

In a thermal reactor with moderators at different temperatures, a difference arises in the average speeds of thermalized neutrons between the high temperature part and the low temperature part of the moderator, and the non-uniformity of the average speed of thermalized neutrons may effect changes in the spatial dependence of the thermal neutron flux in a core. To investigate the thermal neutron flux in the case the average speed of thermalized neutrons is dependent on the position within a core, time-dependent two-group diffusion equations were applied. The influence of a nonuniform moderator temperature on the core power distribution was investigated about a graphite-moderated subcritical reactor driven by periodic injections of pulsed fast neutrons. The cylindrical reactor model by which a high temperature part of a core that has a spallation target at the center is enclosed by a low temperature part of a core was used. Changes in the core power distribution were calculated. It turned out that the momentary increases of a thermal power density caused by periodic injections of pulsed fast neutrons increase as the difference in the average speeds of thermalized neutrons in the high temperature part and the low temperature part of a core increases.     

Determination of critical mass and neutron flux distribution by means of physical models

The design of nuclear reactors, especially new reactors, requires experimental measurements in order to obtain accurate values of the pertinent parameters. In the present paper we present a new method for the preliminary determination of the critical mass of a reactor and the neutron flux distribution; this method is based on the use of physical models. In carrying out these experiments use is made of a model of the reactor which does not contain fissionable material. The working channels in the model are filled with a neutron absorber whose cross section simulates the absorption cross section for neutrons in the fissionable material. The production of fast fission neutrons is simulated by means of a neutron source which is moved along the channels. The distribution of thermal neutrons is measured by means of detectors which are sensitive to thermal neutrons. If the source strength and the absolute value of the neutron flux are known, it is possible to find the critical mass of the reactor.This method has been checked in a reactor with uranium hexafluoride. The value of the critical mass found experimentally was found to be in good agreement with the value obtained when the reactor was started up.The proposed method can also be useful in preliminary investigations of reactor designs, the choice of optimum lattice parameters, etc. The technique is extremely simple and does not require fissionable material or high neutron fluxes.     

Parametric tests and measurements after shimming of a beryllium reflector in a miniature neutron source reactor (MNSR)

After 10 years operation of Pakistan research reactor-2 (PARR-2), a miniature neutron source reactor (MNSR), a beryllium reflector was added to compensate the loss of reactivity due to burn up of fuel. Beryllium shim plates have been placed at the top of the core in a tray provided for this purpose. The control rod was dismantled and withdrawn from the core and the reactor was made subcritical with cadmium shimming. To monitor the neutron population during this experiment, two additional neutron monitoring channels based on BF₃ were installed around the core. Measurement of important Parameters such as effective delayed neutron fraction, decay constant, excess reactivity, control rod worth, temperature coefficient of reactivity, thermal neutron flux, cadmium ratio was done after the addition of Be reflector. Increase in reactivity worth due to addition of Be shim was 1.0 mk.     

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly – a physical model of the SM reactor – and directly in experiments in the reactor. The error is 4.2–10% in determining the reactivity parameters and 5–10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined.     

The neutron tissue dose

This paper is devoted to the determination of dosimetric characteristics of intermediate-energy neutrons. A calculation of the interaction of neutrons with paraffin for the normal incidence of an extended beam of neutrons in the energy interval from 100 ev to 1 Mev, which has been performed by means of an electronic computer, is given here. The computation results are used for calculating the components of the neutron tissue dose resulting from the moderation process in the tissue. The dose resulting from the neutron capture is calculated on the basis of the authors' data on the distribution of slow neutrons in paraffin and on neutron reflection from the surface of a hydrogenous medium. The depth distributions of neutron dose components for thermal neutrons and neutrons with energies of 100 ev, 1, 30, 240, 500 kev, and 1 Mev were obtained. The depth distributions of the biological dose for neutrons of the same energy and the biological dose values pertaining to the unit neutron flux were obtained.