Graphite-moderated Reactor Research Articles

Effect of coolant channel diameter on homogeneous and heterogeneous methods of reactor analysis in reflected subcritical assembly

The systematic variation of neutron flux distribution and buckling measurements as a function of fuel loading in a subcritical assembly (graphite, natural uranium) for two different coolant channel diameters have been investigated. The void ratio (volume of voids/lattice volume) in the two cases was 19% and 24%. The experimental results have been compared with theoretical predictions based on diffusion theory. For full assembly cases the results are in good agreement (as might be expected) with the predictions of Syrett's model for graphite moderated reactors. The correlation of 10 partially filled assembly cases was carried out in three ways: Syrett method lattice parameters with reflector savings calculated from one- and two-group theory for an infinite-slab system.Numerical solution of the two-group, two-region diffusion equations.Two-group heterogeneous (source sink) theory for finite systems. The results show that method (i) breaks down for all cases except those with the full assembly, method (ii) breaks down in nearly all cases and method (iii) holds well down to 36 fuel channels. The breakdown in method (i) and (ii) is most probably due to inaccurate streaming corrections to reactor systems of smaller size. In method (ii), excessive computer time is an additional complicating factor. If the experiments are analysed by heterogeneous methods it is possible to obtain accurate reactor para meters with a partially loaded exponential assembly in cases where only a limited amount of fuel is available.

Process of Homogenization and the Lattice-Parameters of a Sub-Critical Assembly

The systematic variation of lattice parameters as a function of fuel-moderator ratio in a sub-critical assembly (natural uranium, graphite moderated) has been investigated. Diffusion theory was used for correlation of results. The theoretical results were calculated according to the procedure laid down for graphite-moderated reactors based on the unit-cell concept of Wigner-Seitz. The results show that the method (Wigner-Seitz) predicts accurate results even when a fuel element is assumed to be spread over two lattices uniformly. This also proves the method of replacing a square lattice by an equivalent cylindrical one. The effect of voids (gas channels) on the lattice-parameters was also considered.

Process of Homogenization and the Lattice-Parameters of a Sub-Critical Assembly

The systematic variation of lattice parameters as a function of fuel-moderator ratio in a sub-critical assembly (natural uranium, graphite moderated) has been investigated. Diffusion theory was used for correlation of results. The theoretical results were calculated according to the procedure laid down for graphite-moderated reactors based on the unit-cell concept of Wigner-Seitz. The results show that the method (Wigner-Seitz) predicts accurate results even when a fuel element is assumed to be spread over two lattices uniformly. This also proves the method of replacing a square lattice by an equivalent cylindrical one. The effect of voids (gas channels) on the lattice-parameters was also considered.

Cellular changes during malnutrition

tle iatrogenic exposure of the subject to radioactivity, it would be preferable to have a general method for labeling such proteins which would not expose patients (especially infants and pregnant women) to even this small amount of radiation. In the present study, such a method has been devised. H u m a n serum albumin and yGglobulin were labeled with stable bromine by direct halogenation and these proteins were assayed in vitro by activation of the stable bromine with thermal neutrons from a graphite moderated reactor. With this technique, the metabolism of stable brominated h u m a n serum albumin and yG-globulin was studied initially in the mouse and then in man and compared to the metabolism of human serum albumin and yG-globulin labeled with either I TM, 1125, or BrSK It was found that the turnover of the stable brominated human serum albumin and yG-gIobulin as determined by neutron activation analysis was similar to the turnover of their radioiodinated and radiobrominated analogues. The data indicate the feasibility of using such stable tracers in diagnosis and in experimental studies of plasma protein metabolism, such as those currently being used in conjunction with intrauterine fetal transfusions to study the transplacental transfer of plasma proteins from the fetus to the mother.

Resonance Capture and Doppler Coefficient in Cylindrical Uranium Fuel Elements

Double P/0 diffusion theory is shown to be a sufficiently accurate representation for calculating resonance absorption and its temperature coefficient. The theory is formulated to allow for non-uniform temperature distributions and spatial variation of neutron cross sections. It is applied to uranium rods in graphite-moderated reactors, assuming a parabolic fuel-temperature distribution. Volume and surface temperature coefficients for absorption are defined. The energy distributions of these coefficients in strongly absorbing resonances are shown to differ significantly. It is found that the total volume coefficient exceeds the total surface coefficient by 15% at normal operating temperatures. At higher temperatures the total volume coefficient is larger by 5%. Rowlands' formula for the effective uniform temperature is shown to be reliable for calculating the resonance integral and the volume temperature coefficient, but not for the surface coefficient.

Mass-Fission-Yield Curve for Americium-241

Mass-yield measurements were made for the fission of 241Am induced by graphite-moderated reactor neutrons. In most cases gamma-ray spectrometry was used to measure the radioactivity of the isolated fission products. The relative abundance of each nuclide representing a particular fission chain was determined radiochemically with respect to 139Ba, whose fission yield was assumed to be 6.22%. The composite results of the three separate experiments are presented.

A Study of Plutonium Recycle in a Gas-Cooled, Graphite-Moderated Reactor, (III)

Reactor physics characteristics related to plutonium recycling in a gas-cooled, graphite-moderated reactor were investigated and correlated to those for the case without plutonium recycling, using the fuel cycle analysis code FUELMOVE. It was found that, in the initial stages of irradiation, plutonium recycled fuel shows a thermal spectrum harder with respect to neutron temperature and a higher slowing down density, but that differences become smaller as the burnup progresses. A notable difference between the cases with and without recycling Hes in the resonance fission rate which proves to be roughly 20 to 35% higher in the former case.

A Study of Plutonium Recycle in a Gas-Cooled, Graphite-Moderated Reactor, (III)

A Study of Plutonium Recycle in a Gas-Cooled, Graphite-Moderated Reactor, (III)

Effect of massive neutron exposure on the distortion of reactor graphite

Distortion of reactor-grade graphites was studied at varying neutron exposures ranging up to 16×10 21 neutrons/cm 2 (nvt) † † All neutron exposures refer to the number of neutrons with energies greater than 0.18 MeV. at temperatures of irradiation ranging from 425 to 800°C. This exposure level corresponds to approximately 100,000 megawatt days per adjacent ton of fuel MWd/At in a graphite-moderated reactor. A conventional-coke graphite, CSF, and two needle-coke graphites, NC-7 and NC-8, were studied. At all temperatures of irradiation the contraction rate of these graphites increased with increasing neutron exposure in the direction parallel to the extrusion axis, and the needle-coke and CSF graphites contracted approximately the same amount. In the transverse direction the graphites contracted at all irradiation temperatures with the exception of NC-7 at 800°C which contracted and then expanded. Volume contractions were derived from the linear contractions.

Primary Damage State in Neutron-Irradiated Iron

An attempt was made to characterize the primary damage state produced in finite bcc iron samples by neutron irradiation at 0°K, in a series of computer studies. Distributed neutron spectra appropriate to graphite-moderated reactors, water-moderated reactors, and fission neutrons, together with monoenergetic neutron sources with energies up to 10 MeV, were used. The number of primary knock-on atoms displaced by neutron collisions and their energy spectrum were computed using the Monte Carlo method. Anisotropic neutron scattering as a function of energy was described using tabulated experimental data. Damage states produced by these primary knock-on atoms, as a function of their energy, were computed by tracing out elastic atomic collision cascades in the bcc iron lattice. A damage state was described in terms of the displacement spike volume distribution produced, the total number of displaced atoms, and the size distribution of vacancy and interstitial atom clusters in a displacement spike. A description of how the damage state depends upon sample geometry and shape, the incident neutron energy spectrum, and the incident neutron angular distribution is given. It is shown that the damage state produced in samples of different size and shape can be significantly different and the irradiation conditions required to produce equivalent damage states in different sample sizes are defined. The principal content of the article pertains to the primary basis for the change in mechanical properties caused by neutron irradiation, as indicated by current theories.

The thermal expansion of polycrystalline carbon and graphite from −196°C TO 2000°C

The thermal expansion of artificial graphite is of particular interest to designers of high-temperature graphite-moderated reactors since it correlates with the dimensional changes which occur under fast neutron irradiation. Measurements have been made of the thermal expansion coefficients of a number of different artificial polycrystalline graphites over the temperature range −196°C to 2000°C. It is shown that, above room temperature, the thermal expansion coefficients of all the graphites measured rise linearly with temperature. An analysis of the results, made in terms of the single-crystal expansion, indicates that changes in artefact structure above 800°C correspond with analagous changes resulting from neutron irradiation.

Neutron Thermalization in Graphite

In this work our ability to understand thermal-neutron spectra in graphite-moderated reactor systems is examined on the basis of a theoretical description of neutron scattering that begins at the microscopic level.The first step in this examination consists in determining the extent to which current ideas of lattice vibrations in graphite are consistent with measurements of the scattering law and of specific heats. Theoretical scattering laws and specific heats based on a few different models for lattice vibrations are compared with experimental results. The theoretical scattering law is calculated within the framework of the incoherent and Gaussian approximations. The question of the accuracy of the latter approximation is discussed in detail. No estimates have been made of the magnitude of the uncertainty introduced by the use of the incoherent approximation.Following the discussion of neutron scattering at the microscopic level, we show (1) the sensitivity of various integral properties of the scattering kernel and of thermal-neutron spectra in a homogeneous medium to the frequency distribution of lattice vibrations, and (2) the accuracy of the Gaussian approximation for use in computing thermal-neutron spectra in graphite. Finally, a detailed theoretical model for the scattering of neutrons by graphite is applied to the problem of comparing calculated neutron spectra with the measured spectra in two strongly heterogeneous, graphite-moderated assemblies.These considerations show that current theoretical ideas concerning the frequency distribution of lattice vibrations in graphite are consistent with the results of the measurement of the scattering law, with the specific heat, and with most of the available results of the measurements of thermal-neutron spectra in reactor-like configurations.

Effect of water content in concrete on thickness and cost of reactor shielding

Results of biological shielding calculations for various types of concrete are compared in this paper. A graphite-moderated and graphite-reflector reactor is considered. The thicknesses and costs of biological shielding using concretes of different types and densities is analyzed as a function of the water content in the concretes. The goal of the authors is to determine, on the basis of the 46 most commonly used types of concrete in reactor construction throughout the USSR, the economic feasibility of increasing the amount of water contained in concrete

Hallam Nuclear Power Facility Electric Auxiliary-Power Systems

This paper describes the electric auxiliary-power systems in a sodium-cooled graphite-moderated nuclear reactor (SGR). Reliability requirements and methods of achieving them are discussed, and sodium-system preheating, a requirement unique to sodium-cooled reactors, is outlined. Nuclear instrumentation electric noise problems and methods of avoiding them are included. Auxiliary power requirements of a 360-mwe (electrical megawatt) SGR plant are compared with those of a similar conventional plant.

Effets de l'irradiation neutronique sur la conductibilite thermique des graphites nucleaires

Two instruments, both briefly described, are used to measure the thermal conductivity of irradiated graphites. One uses the principles of heat transfer and permits the thermal conductivity to be measured near room temperature. The other relies on the method due to Kohlrausch and provides a measure of the conductivity as a function of temperature. Samples taken from the body of graphite-moderated reactors have been examined. From the measurements, the variation of ( K o K i ) − 1 with the dose has been determined for various irradiation temperatures. K o and K i are thermal conductivities before and after irradiation, respectively. At the same time, curves giving, for various doses, the variation of ( K o K i ) − 1 with irradiation temperature have been determined. These results agree satisfactorily with those published by other investigators. The variation of the thermal conductivity with the temperature of measurement was studied on samples of reactor-grade graphite irradiated at various temperatures in a heavy-water reactor. Measurements were carried out on each sample from ambient temperatures up to the temperature at which it had been irradiated. From the results it was possible to deduce, for each type of graphite investigated, the variation of ( K o K i ) − 1 as a function of the irradiation temperature and the temperature of measurement.

黒鉛減速ガス冷却炉におけるプルトニウムリサイクルの研究, (I)

The IBM 704, 709 or 7090 computer code developed at the M. I. T. has been used to analyze the fuel-cycle performance of a gascooled, graphite-moderated reactor. For the first time this code has been used to treat plutonium recycle study in a thermal reactor. Fuel feed cases studied consist of natural uranium, 1.0 _??_ or 1.3_??_ enriched uranium blended with recycle of all, 90%, 75% or none of the plutonium produced in the reactor. The two methods of managing fuel which were studied were simple batch irradiation and bidirectional fuel management.Use of bidirectional fuel management and plutonium recycle greatly increases the amount of energy that can be liberated from a given amount of uranium fuel of specified enrichment. For example, the number of MWd of energy that can be obtained from a metric ton of 1.0 _??_ enriched uranium fuel with various methods of fueling are:Batch irradiation, no recycle 4, 616Batch irradiation, total recycle 7, 660Bidirectional irradiation, no recycle 11, 352Bidirectional irradiation, total recycle 15, 191 Even with natural uranium feed, bidirectional fueling and total recycle of plutonium permits production of 9, 274MWd/t. of uranium.The high burn-up attainable with plutonium recycle is advantageous also in flattening the power density distribution in the core. This permits operation of the reactor at a higher total power when steady state plutonium recycle conditions have been reached than when the reactor is first charged with fuel of uniform composition.Poisoning due to the build-up of 242Pu is less severe than had been expected and could be controlled by-bleeding off and not recycling 1025% of the plutonium.

The oxidation of magnesium alloys in reactor atmospheres

Detailed results are given of the oxidation behaviour of several magnesium-base alloys in normal reactor atmospheres and some “fault conditions”. The alloys A. 12 (Mg-Al-Be), ZA (Mg-Zr), ZW. 1 (Mg-Zr-Zn) and AM. 503 (s) (Mg-Mn), chosen because of their possible application as canning or structural materials in a gas cooled graphite moderated reactor, oxidise similarly with the production of thin adherent films under normal reactor gas atmospheres, the oxidation rates being very low even at 600° C. The effects of temperature, surface finish, moisture and air content of the CO 2, welding and stress on the oxidation of these alloys in carbon monoxide/carbon dioxide mixtures have been studied in the temperature range 500–700° C. The possible reactions contributing to the oxidation process are discussed.

L'utilisation de l'alliage Mg-Zr comme materiau de gainage

The authors present the properties of the Mg/0.6 wt % Zr alloy, considered from the point of view of its use as canning material for fuel elements in CO 2-cooled graphite-moderated reactors. After describing the various methods of fabrication (wire-drawing, machining, etc.) and of canning, the authors show how the practical requirements can be satisfied by the principal properties of this alloy, viz. mechanical strength, ductility, metallurgical stability towards heat treatment, weldability and good stability under irradiation.

Symposium on the Physics of Graphite Moderated Reactors, Bournemouth, April 1962

Symposium on the Physics of Graphite Moderated Reactors, Bournemouth, April 1962

Thermal Properties of Air-Cooled Graphite Channels

A summary is given of the work done on the thermal behavior of air-cooled graphite channels. The differences in results between this work and work done by Robinson and Taylor (5) appear to arise from the different models used. The results reported here and re-evaluation of the Robinson and Taylor work indicate that their model is not confirmed by their experiments and does not predict the limit in permissible temperatures imposed by graphite oxidation in air-cooled graphite moderated reactors.