Summary: An EVESR MKI prototype fuel bundle was fully instrumented and operated intermittently for a 5-month period at the Pacific Gas and Electric Company’s Moss Landing Power Station. The vessel was operated up to 1000 psi with steam flows from 3000 to 26,600 lb/h, and steam inlet temperatures up to 825 degrees F. Data was recorded for blowout, vibration, flow distribution, heat transfer and pressure drop. The mechanical integrity of the fuel bundle, riser, and jumper system was satisfactory and considered to be of adequate design. No significant vibrations were noted during the various phases of operation. Average flow distribution in three of the inner tubes showed an average variation of 5 percent from equal distribution. The center and corner tubes were low and the side tube was high. Maximum deviation, from an equal one, measured 12 percent. Blowout of the flooded fuel bundle was accomplished with dry or significantly wet 1000 psia inlet steam, that steadied out to a minimum flow of 1250 lb/h. Blowout times were estimated at less than a minute for all flows above 1250 lb/h, and times in the vicinity of 2000 lb/h were estimated to be in the order of 5 to 15 seconds. …

From introduction: "This is the eighteenth in a series of quarterly reports which cover the progress and results from the conceptual designs, economic evaluations and research and development work performed by the General Electric Company as part of Contract AT(01-3)-189, Project Agreement No. 13."

Summary: An analytical study of the stability of the Big Rock Nuclear Reactor has been performed for the plant as built, and supplements a previous design stability study. The plant has been determined by this analysis to be very stable under every mode of operation anticipated during Phase I of the development program testing. Even under conservative assumptions of system parameters the minimum calculated gain and phase margins do not go below 13.0 db and 46 degrees, respectively. (Nor are these both reached simultaneously for the same operating condition.) These are characteristics of a very stable, well-behaved system. In addition to this analysis, a second, less conservative series of computations was performed to provide expected realistic closed loop data for comparison with Phase I test results. The most responsive test thus predicted occurs at 60 percent power, 1500 psia, minimum flow, and maximum subcooling. For this case the closed loop peak response of power to reactivity occurs at a frequency of 0.90 cycles per second with an amplitude of 3.90 db. This corresponds to an expected open loop gain margin of 16.5 db and a phase margin of 63 degrees. Although knowledge of reactor transfer function is to be determined …

Abstract; The fission product data obtained during SM-1 Core I operation (June 1957 - May 1960) is reviewed briefly and interpreted. Evidence is presented to indicate that a fuel element defect was responsible for the high fission product activity level observed in the primary coolant. Relative escape coefficients are calculated and the defect size estimated. Anticipated fission product levels during SM-1 Core II and SM-1A Core I operation are estimated from alpha surface contamination data on completed fuel elements. The importance of in-line sampling for monitoring fission product activity is stressed as well as the need for failed fuel element detection methods.

Abstract: Test procedures for special tests involving in-core SM-1 temperature and flow instrumentation are described (Task XIV Package Tests). These tests involve in-core steady state flow and temperature measurements, loss of flow transients, load transients, reduced primary system pressure operations and reduced element flow. The thermal and hydraulic conditions prevailing in these tests, including steady state and transient burnout rations, are developed. The effects of reduced system pressure and flow on the burnout ratios are determined as are the expected stuck rod conditions when Task XIV test elements are installed. The effect on the maximum credible accident is included and a recommendation to conduct these Task XIV package tests is made.

Technical report describing how corrosion data and exposure effects were obtained by subjecting metallic samples, during programmed test runs to flowing sodium in 6 test loops fabricated with various combinations of three selected materials, Type 316 stainless steel, 2 1/4 Cr-1 Mo alloy steel, and 5 Cr-1/2 Mo-1/2 Ti alloy steel. Information produced by each test run, including operational and metallurgical data and analyses, is presented. Data are shown in tables, graphs, and drawings.

Abstract: The natural circulation, direct cycle, boiling water reactor reference design presented in this technical report is the alternate to the preferred preliminary design developed under Phase I of the PL-3 contract. The report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and training program outline.

Abstract: The pressurized water reactor preliminary design presented in this volume is the preferred design developed under Phase I of the PL-3 contact. This technical report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and a training program outline.

Summary: Stainless steel clad nuclear fuel cycle costs can be reduced to those associated with Zircaloy clad fuel or potentially lower by reducing the thickness of the clad tube wall until performance penalties offset the savings associated with the reduction in parasitic neutron absorption. To demonstrate the feasibility and investigate performance capabilities of thin clad fuel rods for power reactor application an assembly was fabricated with 0.0127 cm (5 mil) thick stainless steel cladding tubes for irradiation testing in the Vallecitos Boiling Water Reactor (VBWR). The fuel bundle was placed in the VBWR and irradiation was begun in November, 1961. The irradiation is scheduled to continue until the target exposure of 2.74 x 10(20) fissions/cc (10,000 MWD/T of uranium) average burnup is reached. Destructive examinations of fuel rods will be performed at regular intervals throughout life to determine fuel rod performance.

The extrusion forming of ceramic powders may be economically interesting in the field of nuclear fuel fabrication. When applied to the forming of rod-type uranium dioxide fuel, extrusion processes have been able to produce cylindrical bodies with length-to-diameter ratios much greater than those of the conventional die-pressed pellets. Furthermore, after being sintered, the extrusions have exhibited densities at least as high as those of sintered pellets. Thus, extrusion forming may offer reductions in handling during fabrication and, at the same time, provide a fuel with improved performance characteristics by decreasing the number of discontinuities in the fuel column. This report reviews the production of these extrusions, sets forth some of their characteristics, describes the materials and processes employed in cladding them, and records the pre-irradiation data pertaining to the finished fuel rods and fuel assembly. Irradiation of the fuel assembly in the VBWR was initiated on July 17, 1962.