Nuclear Structural Engineering最新文献

The high temperature difference between exterior and interior cladding in an annular combined boiling-superheat fuel element results in considerable axial stresses and strains in both claddings during reactor operation. Each start-up or shut-down of the superheat reactor causes cyclic changes of these stresses and strains.

In the first part of this report the equations necessary to predict the behaviour of boiling-superheat fuel elements are deduced and the resulting stresses and strains are given for several examples with various cladding materials.

The second part of this paper reports the results of experimental studies with boiling-superheat fuel elements in an out-pile thermal cycling test facility. Full size fuel element specimens have been subjected to up to 800 severe thermal cycles. The observations made, including some defects which have occurred, are discussed. The results are critically reviewed and compared with the analysis of part one.

Even considering the fact that the mechanical properties under irradiation at high temperature of the cladding materials of interest are not very well known today, it is felt that the axial stresses and strains in a boiling-superheat fuel element do not present an unsurmountable problem in the development of this type of fuel elements.

The present paper describes the dynamic analysis applied to the reactor containment building of the JPDR, which has been completed in 1963. As the method of evaluating the most probable earthquake response for aseismic design was explained in the previous paper of part I, the analysis included in this paper is restricted to the calculation of the vibrational properties of the building, performed on the basis of an assumption that it undergoes a rocking vibration. In addition, the measurement of vibrations of the building during small earthquakes is reported and some data are presented as a result of the spectral analysis. As measurements are continuing, only a tentative conclusion as regards the natural period is made by comparison between calculated and measured results.

Detailed neutron flux measurements have been performed in a 48 cm thick configuration of thin regions of Fe and D₂O (30% of total thickness Fe), both when the Fe regions are massive and when they are penetrated by 15 cm dia. D₂O channels. It was found that the total leakage of neutrons through the configuration was increased by 25% in the latter case.

It is shown that this increase and the detailed flux distribution on a duct axis can be satisfactorily predicted by a one-dimensional shielding code when the duct is calculated as a pure D₂O layer and a radial buckling term is used for the < 1 eV neutrons when penetrating the Fe regions. Another calculation through the massive part of Fe is to be performed in the usual way. It is believed that this 2-component method is usable in a wider range of similar configurations.

A calculation with regions where the Fe and the D₂O ducts have been homogenized into a single material overestimates the increase in leakage, and the relative error is greater than that in the 2-component calculation.

The transient temperature distribution in bonded end caps of plate-type fuel elements submitted to sudden temperature change resulting from unsteady-state reactor operation is analytically determined. The solution for the temperature distribution is represented by products of circular function, exponential functions and coupling relations between the end caps and the fuel material of the fuel elements. From the calculated results of a numerical example given for a ceramic dispersion fuel element with zircaloy end caps, the following points of primary interest are found:

• 1.

  (a) For a given end-cap material, the temperature distribution for end caps of various depths (or thicknesses) is the same, if the time of heat conducting is proportional to the square of the depth.

• 2.

  (b) The temperature distribution in the end cap decreases with increasing end-cap length and Fourier number.

• 3.

  (c) Repeated sudden temperature changes that induce severe thermal shock, cycling, fatigue, and temperature excess can seriously affect the corrosion rates and impair the structural integrity of the end caps as well as the fuel elements. Further, the corrosion temperature limit and thermal fatigue of the end cap (or cap of a fuel can) can be much more serious than that of the fuel material.

Some ideas and suggestions for the utilization of concrete as a containment structural material are discussed. The purpose of this paper is to show that concrete can be used expediently only if one does not attempt to achieve simulation of individual qualities of steel with it, but if, starting from the properties and potential of this material, appropriate ideas of construction are pursued. In this context two considerations seem to be particularly valuable:

• 1.

  (1) When considering the variation of internal pressure with time, attention should be paid to the utilization of the large heat capacity of concrete. There follows: (a) A more realistic insight in the stress field within the concrete wall can be gained. (b) With the lowering of internal temperature the internal pressure decreases, thus giving rise to rapidly changing leak rates. (The definition of the leak rate gives a period of 24 hours, which, therefore, must be employed accordingly.)

• 2.

  (2) For reasons of construction and of protection against radiation it is often necessary to plan an intermediate space, which - together with the decreasing internal pressure and the large thickness of the wall - affects the overall leak rates.

Slip ratios are presented for air-water, nitrogen-mercury and nitrogen-freon-113 mixtures flowing at superficial liquid velocities up to one foot per second. For non-circulating mixtures, the void fraction is presented as a function of the superficial gas velocity in air-water and in nitrogen-mercury.

To thoroughly study the effects of the major variables in forced-convective burnout, 402 high-pressure critical heat flux determinations were made with seven uniformly heated, internally-cooled, tubular test sections using water, subcooled at the inlet, as the coolant. Data were collected over the following range of variables: $\text{Critical heat flux}\text{0.44 to 2.57 × 10}{}^6\text{Btu/hr-ft}{}^2\text{Mass velocity}\text{0.5 to 13.7 × 10}{}^6\text{lb/hr-ft}{}^2\text{Exit coolant conditions}\text{116.6°F subcooled to 59.2\% quality}\text{Pressure}\text{500 to 1500 psia}\text{Heated length}\text{24 to 77}\text{5}\text{8}\text{in.}\text{Inside diameter}\text{0.245 to 1.475 in.}$

The 1.475 in. tube represents the largest tube ever tested.

The results indicate the existence of three burnout regimes: nucleate boiling, annular flow, and the still not fully explored “transition” region.

An analytical study of annular flow burnout based on observations and dimensional analysis yielded a correlation which relates the critical heat flux to the significant local fluid properties and flow parameters. This equation was used to correlate and study the critical heat flux data obtained in the present work together with a selection of additional data points from the major available sources, including data for freon-12. The equation contains four dimensionless groups and constants determined by regression analysis with a limited portion of the data. Critical heat fluxes calculated by the correlation are in good agreement with the measured values.

Observations of the effects of the parameters in subcooled nucleate boiling burnout enabled the formulation of an empirical equation for this regime. The equation relates the critical heat flux only to the mass flow rate and subcooling. In testing the equation with available data it was found that the correlation holds throughout the 60 to 2000 psia pressure range equally well, thus implying that there is no pressure effect on subcooled burnout over this entire range.

The complete form of this study, including tabulated data, correlating procedures, literature survey, etc., is available as ref. [1] of the bibliography.

In erecting the prestressed concrete pressure vessels used for natural uranium, graphite moderated, gas cooled reactors, the designer is faced with the important difficulty of the thermal insulation for the prestressed concrete vessel. For its power plants, E.D.F. 3 at Chinon and E.D.F. 4 at Saint-Laurent-des-Eaux, Electricité de France has used an insulating material, specially studied and developed for this purpose: pumice concrete. A brief recapitulation of the properties needed for insulating the reactor pressure vessels is given here, followed by detailed information on the nature and the source of the pumice concrete components, the process and manufacture checkings and its use in the reactor, and indications concerning principal mechanical and thermal properties. Finally, instructions are given on how to put the pumice concrete to various uses.

Discussion of the principles of radiation shielding design for cobalt-60 teletherapy installations. The requirements for radiation protection are outlined, and the influences of general layout and operational aspects on the design of the shielding for teletherapy rooms are indicated. Shielding design examples are given.