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Data from top injection ECCS tests conducted at Columbia University have been analyzed as part of an effort to qualify the FLASH-6 computer program for performing post-blowdown heat transfer calculations for the LWBR Safety Analysis. These experiments, which employed a full-scale fuel assembly with electrical heater rods to simulate an inlet rupture for a pressurized water reactor, provided test conditions and rod cooling mechanisms quite similar to those encountered in the postulated LWBR cold leg break loss-of-coolant accident. Clad temperature predictions were obtained using both the modified Dittus-Boelter and the Dougall-Rohsenow correlations to evaluate beyond CHF heat transfer coefficients. Overall comparisons using the FLASH calculated flow rates indicated that the rod temperature calculations were conservative using either of the heat transfer correlations because virtually none of the coolant was calculated to penetrate the heated test assembly. Heat transfer model comparisons were also performed by adjusting the calculation so that coolant was injected directly into the top of the rod bundle to simulate the experimentally observed flow conditions. Once this downflow was forced, conservative temperature predictions were obtained using the Dougall-Rohsenow correlation, whereas the modified Dittus-Boelter beyond CHF option yielded non-conservative results.

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

As part of the Light Water Breeder Reactor development program, hydraulic tests of reactor components were performed. This report presents the results of several of those tests performed for components which are somewhat unique in their application to a pressurized water reactor design. The components tested include: triplate orifices used for flow distribution purposes, multiventuri type flowmeters, tight lattice triangular pitch rod support grids, fuel rod end support plates, and the balance piston which is a major component of the movable fuel balancing system. Test results include component pressure loss coefficients, flowmeter coefficients and fuel rod region pressure drop characteristics.

The PDQ-8 program is designed to solve the neutron diffusion, depletion problem in one, two, or three dimensions on the CDC-6600 and CDC-7600 computers. The three dimensional spatial calculation may be either explicit or discontinuous trial function synthesis. Up to five lethargy groups are permitted. The fast group treatment may be simplified P(3), and the thermal neutrons may be represented by a single group or a pair of overlapping groups. Adjoint, fixed source, one iteration, additive fixed source, eigenvalue, and boundary value calculations may be performed. The HARMONY system is used for cross section variation and generalized depletion chain solutions. The depletion is a combination gross block depletion for all nuclides as well as a fine block depletion for a specified subset of the nuclides. The geometries available include rectangular, cylindrical, spherical, hexagonal, and a very general quadrilateral geometry with diagonal interfaces. All geometries allow variable mesh in all dimensions. Various control searches as well as temperature and xenon feedbacks are provided. The synthesis spatial solution time is dependent on the number of trial functions used and the number of gross blocks. The PDQ-8 program is used at Bettis on a production basis for solving diffusion--depletion problems. The report describes …

The DECAG (Deformation of Cladding into Axial Gaps) ex-reactor test program evaluated deformation of Zircaloy-4 cladding into axial gaps in tubular fuel elements. These axial gaps are the result of cladding elongation and fuel stack shrinkage. The test program consisted of twelve series and subseries of both fully recrystallized and stress-relieved highly cold worked tubing tested under pressure-temperature combinations in autoclaves. The test program also verified the validity of achieving test acceleration through the use of elevated temperatures by correlating both ovality and diameter change at lower temperatures with the Larson--Miller Parameter.