A.R. Wazzan

Thermal-hydraulic characteristics of pressurized water reactors during commercial operation: IV. Saturation pressure, circulation ratio and carry-under in a PWR steam generator (Tricastin 1 nuclear power plant)

Abstract Saturation pressure, downcomer velocity distribution (angular), local circulation ratio (angular distribution), overall circulation ratio, and carryunder are measured as a function of load in the steam generator No. 20 of the PWR Tricastin 1. Carryunder is found to be negligible; It reaches a maximum value of about 0.8% at full load. The circulation ratio decreases with load; it decreases from 22 at 11% of full load to 4.4 at full load. At full load the measured circulation ratio, 4.4, compares with a design value of 4.1. Variation of the circulation ratio and saturation pressure with load reported here for Tricastin 1 are in good agreement with results reported earlier for the PWR Bugey 4.

I. Circulation ratio and carry-under in a PWR steam generator (Bugey-4 nuclear power plant)

Abstract Thermohydraulic measurements made on steam generator No. 14 (Westinghouse type 51A) of unit 4 of Bugey (a French PWR) allowed the determination of the circulation ratio and the percentage of carry-under as a function of the percentage of full load. At full load the circulation ratio, obtained from a mass balance, was found to be about 4.6, rising to about 14.6 at 30% of full load. Both exceed the estimated design value of 4.17 at full load. The percentage of carry-under is found to be negligible and relatively independent of load up to about 50% of full load, but increases thereafter to about 1.23% at full load. This level of carry-under does not appear to present any cavitation problems in case of sudden depressurization.

Computer modeling of steady state fission gas behavior in carbide fuels

Abstract A model for the simulation of long-term, steady-state fission gas behavior in carbide fuels is formulated. It is assumed that fission gas release occurs entirely through gas atom diffusion to grain boundaries and cracks. Fission gas bubbles are assumed to remain stationary and to grow as the net result of gas atom precipitation into the bubbles from the matrix solid and gas atom re-solution from the bubbles into the matrix. Furthermore, assuming that local gas atom redistribution process in the immediate neighborhood of a bubble is very rapid, the bubble size is assumed to correspond to the equilibrium size that maintains exact balance between the rate of gas atom re-solution and that of gas atom precipitation. The model also treats the effect of attachment between bubbles and second-phase precipitates; the experimentally observed faster growth rate of precipitate bubbles is simulated using a reduced re-solution parameter for precipitate bubbles. With the grain matrix assumed to be spherical, the model allows the computation of the radial distribution of the intragranular bubbles and the gas atom concentration in the matrix. The flux of gas atoms arriving at the grain boundary is computed. The continual growth of grain boundary bubbles, resulting from the accumulation of gas atoms on the grain boundary, leads to grain boundary interlinkage and all gas atoms that subsequently reach the grain boundary are assumed to be released. Similarly, all gas atoms generated following the interlinkage of intragranular bubbles are also assumed to be immediately released. Application of the model indicates that fission gas swelling is largely due to intragranular bubbles. Grain boundary bubbles, although very large in size, contribute little to fission gas swelling and the contribution from gas atoms in solid solution in the matrix is even less significant. Physical parameters entering the model were assigned numerical values that closely represent the physical characteristics of the irradiation samples. Careful comparisons between the results of sensitivity studies and the experimental data readily identify the re-solution parameter to have the strongest influence on the results predicted by the code and that the grain size, and not the temperature, is the dominant factor affecting gas release. When allowance is made for the uncertainties of the experimental data, the predicted fission gas swelling also correlates well with experiment. The spread in the fuel swelling data, however, indicates that fuel cracking, and not fission gas swelling alone, very often contributes significantly to the fuel external dimensional changes. The linear fission gas swelling rate prediceted by the model exhibits almost a linear variation with temperature. This result correlates well with the linear swelling rate obtained from experimental swelling data if immersion density data alone are used, in order to eliminate the sources of uncertainties associated with fuel cracking.

Simulation of non-equilibrium fission gas behaviour during fast thermal transients☆

A code for predicting the behavior of non-equilibrium fission gas in oxide fuel elements undergoing fast thermal transients is developed. A new variable, the equilibrium variable (EV), is introduced which, together with bubble radius r, completely specifies a fission gas bubble with respect to its size and equilibrium condition. The code is used to simulate the measurements in two TREAT transients with peak temperatures of 2477 and 2000 K. The computations are in fair agreement with the observations for bubbles smaller than 964 A in diameter, but not for the larger bubbles. In all simulations, bubbles that grew during the heat-up phase of the transient were found to be “frozen” at a larger than equilibrium size during the cooldown phase of the transient. This phenomenon can significantly affect posttransient swelling and gas release. It is also found that the assumption of equilibrium can introduce considerable error in the computed bubble distribution, swelling and gas release at the end of as well as at post fast thermal transients; for example, the non-equilibrium model releases more gas. The code is also used to simulate the H3 TREAT transient as analyzed by Stahl and Patrician (initial temperature equal to 785 K with a maximum of 2393 K attained in 4.2 seconds, maximum thermal gradient of 10 000 K/cm and grain diameter of 4 to 10 μm) using the ideal gas as well as the Van der Waals equations of states. The gas inventory at the start of the transient is assumed to be at equilibrium in the smallest radius group (6.2 A), and the initial bubble concentration is assumed to be 1.2 × 1019/cc. Release rate is found to be strongly dependent upon grain size and initial bubble concentration; a 4 micron diameter grain releases about 95% of the gas retained at the start of the transient, while 6 and 10 micron grains release 68% and 20% respectively. When the initial bubble concentration is reduced by a factor of 16 for the 10 micron grain, fractional release increases to 62%. Gas release is found to result primarily from small bubbles (r < 20 A).

Steady-state fission gas behavior in uraniumplutoniumzirconium metal fuel elements

Abstract An analysis of fission gas release and induced swelling in steady state irradiated U—Pu—Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior.

Thermal-hydraulic characteristics of pressurized water reactors during commercial operation: VI. Chemical analysis of secondary water in the steam generators of Bugey 4 and Tricastin 1

Abstract Chemical analyses of the secondary water in the steam generators of the PWRs Bugey 4 and Tricastin 1 are given. Turbidity (in equivalent X ppm SiO 2 ) distribution across the tube sheet and its variation with load is discussed in terms of the feedwater distribution onto the cold and hot legs. Sludge/magnetite distribution onto the tube sheet is discussed in terms of the thermal hydraulic of the steam generators. Two methods are given for estimating, analytically, the weight of this deposit.

III. PWR steam generator response to a scram at 50% load and an open grid transient (bugey-4 nuclear power plant)

Abstract Two transients, an open grid and a scram at 50% load, were conducted on unit 4 of the PWR power plant Bugey. The thermal hydraulic response of the steam generator was recorded. For the open grid test, the following observations are noted: No alarming phenomena are observed in the steam generator during the transient. Primary pressure oscillations were very mild, and did not exceed about 4.8 bar/min with a maximum amplitude of ±8 bar. This condition should not result in significant stress levels. Steam generator outer shell metal temperature gradients remained within very acceptable limits; a maximum amplitude of about +13°C and a rate not exceeding 0.8°C/min are obtained. This slow rate is explained by a fall in primary water temperature that allows for a temperature decrease inside the U-tube bundle. Similarly, the temperature rise on the tube sheet does not exceed an amplitude of 20°C with a rate of about 2°C/min. Again these conditions do not lead to any significant thermal shock on the tube sheet. The steam generator feed controls maintain the level within the normal operation range and the small addition of colder feedwater does not lead to great temperature changes because of the large mass of the recirculation water in the steam generator. For the scram at 50% load, the following observations are noted: no severe thermal or pressure transients are observed in this test. Fluid temperature fluctuations occur with rates not exceeding 1°C/s and a maximum amplitude of about 20°C in the downcomer and 10°C on the tube sheet. Steam generator outer shell temperature varies at a rate of about ±0.8°C/min with a maximum amplitude of about 16°C. These thermal transients should lead to thermally induced stresses of acceptable levels.

II. Steady state thermal measurements on the secondary side of a PWR steam generator (BUGEY-4 nuclear power plant)

Abstract Steady state thermal measurements, water and shell temperature, are made in the downcomer of the steam generator of the PWR BUGEY-4. The measurements show that the flow is of a swirling type with the degree of swirl being a function of the load (power). The angular temperature distribution of water and shell exhibits a minimum that rotates from the hot leg to the cold leg with increasing load. The results also admit the conclusion that the feedwater and recirculation water are well mixed in the cylindrical section of the downcomer. Measurements and computations of fluid and shell temperatures are in good agreement for the steam generator in a state of natural circulation. For the state of forced circulation, the agreement is poor when it is assumed that the flow in the downcomer is turbulent and the steam generator outer shell is well insulated. The agreement is excellent for turbulent flow but with air leakage, or infiltration, between the steam generator outer shell and its insulation. If the hypothesis of air infiltration is ruled out, agreement between measurements and computations is much improved when the flow in the downcomer is taken as a turbulent core flow with an attendant laminar boundary layer. The existence of a laminar boundary layer would require the flat plate transition Reynolds number be enhanced by a factor of 100. This enhancement could result from the combined effects of swirl and density gradient in the downcomer.

Thermal-hydraulic characteristics of pressurized water reactors during commercial operation: V. Steady state thermal measurements on the secondary side of a PWR steam generator (Tricastin 1 nuclear power plant)

Abstract Measured downcomer water and metal shell temperatures in the steam generator No. 20 of the PWR Tricastin 1 show that the downcomer flow is of the swirling type, just as found previously in Bugey 4. A comparison of results for Tricastin 1 and Bugey 4 shows that the addition in Tricastin 1 of a flow distribution baffle plate, between the tube sheet and the first cross plate, while reducing the height of the opening between the tube sheet and the shell surrounding the bundle, may have resulted in the observed reduction (by a factor one half) of sludge deposit upon the tube sheet in Tricastin 1, and in fixing, with extended period of operation, the boiling zone in the cold leg (a desired event) and near the tube-free tube lane.

Thermal-hydraulic characteristics of pressurized water reactors during commercial operation. VII: Thermal-hydraulics of the PWR Paluel 1 steam generator during commercial operation

Downcomer fluid velocity, shell, fluid, and tube sheet temperature, and feedwater distributions across the cold and hot legs of Paluel 1 steam generator No. 81 are measured as a function of load. The results indicate that swirling motion in the downcomer is negligible compared with results reported earlier for the PWRs Bugey 4 and Tricastin 1. Overall circulation ratio and saturation pressure decrease almost linearly with increasing load. Carry- under and carryover are found negligible at all loads. A nonuniformly drilled feedwater ring distributes 80% of the feedwater onto the hot leg and 20% onto the cold leg. The combination of hydrocyclonic steam-water separator dryer design and nonuniformly drilled feedwater ring is successful in retaining a high percentage (∼ 70%) of the rather pure (compared with recirculation water) feedwater at tube sheet level across the hot leg, even at full load. Consequently, chemical quality, and turbidity level of secondary water above the tube sheet in the hot leg is found superior to that in the cold leg.