Introduction: The objective of Project Agreement 22 is to determine the feasibility of using in-core ion chambers to cover the complete reactor neutron flux startup range from 10(4) -5 - 10(13) nv using in-core ion chambers. The counting mode of operation will be used at low neutron flux levels and the RMS voltage fluctuation mode (Campbell Theorem) will be used at high neutron flux levels. The June-September Progress Report (GEAP-4386) shows how the RMS voltage mode can be used, discusses counting problems with long cable and ways of maximizing signal levels. This report discusses primarily the effect of gamma on counting with in-core ion chambers and the range of neutron flux measurable in the RMS voltage mode. Readers are referred to GEAP-4386 for a summary of all previous progress to attain the objective of PA-22.

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 …

Technical report describing that voids were measured in a ½-inch by 1-3/4-inch channel with the S-1 insert (B(0)/B(1) = 0.4, L(0) = 0.1 inch), at 2 inches ahead of the insert (position A), ½-inch past the insert (position B), 5 inches past (position C), and 12 inches past (position D). The conditions were: P – 1000 psia, G = 1.00 x 10(6) lb/h-ft(2), and x = 18.8 percent. Average void and void distribution at position A are the same as for flow in a straight channel. Void distribution at position B shows that the stagnation region downstream of the inserts contains a high fraction of voids. Average void and void distribution at positions C and d show that the two-phase mixture becomes strongly mixed (homogenized) as a result of passing through he contraction-expansion inserts. Distribution at position D approaches the distribution at position A; i.e., the straight channel distribution.

Development of nuclear reactor cores having high power density, long fuel life, and low fabrication costs is the objective of this program sponsored by the AEC. Five tasks are in progress: (1) Task 1A-High Power Density Fuel Development, (2) Task 1B-Fuel Fabrication Development. Assembly, (3) Task II-Stability, Heat Transfer and Fluid Flow, (4) Task III-Physics Development, and (5) Task IV-Co-Ordination and Test Planning.

The objective of Project Agreement 22 is to determine the feasibility of using in-core ion chambers to cover the complete reactor neutron flux startup range from 10(4) -5 - 10(13) nv using in-core ion chambers. This technical report discusses the following topics: low versus high cable termination impedance, amplifier considerations, noise considerations, gas and pressure selection, cable selection, effect of gamma, effect of temperature, and remaining problems.

Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter …

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.

The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.

Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. In the course of this program, a new theory was suggested and an experimental apparatus was designed and built. Experiments were carried out to test the new model. This present report contains additional data and information extracted from the experiments at PG&E Humboldt Bay Power Reactor at Eureka, California. During the last days of 1963 a number of control rod and fuel bundle worth measurements were made in the ESADA Vallecitos Experimental Superheat Reactor (EVESR) using the (k[beta]/[script l] technique. A description of the experiments is given in the text of the report and some results are reported. A computer program was written to perform the data analysis of the pulsed neutron experiments and the code is discussed in the Appendix.

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.

Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities.

Development of nuclear reactor cores having high power density, long fuel life, and low fabrication costs is the objective of this program sponsored by the AEC. Five tasks are in progress: (1) Task 1A-High Power Density Fuel Development. All fuel irradiation has been terminated with the final shutdown of the VBWR. The high burnup average achieved by a single assembly in the group is 10,000 MWD/T (assembly 1F). Twenty-one of the original 24 assemblies have failed or are suspected of failure. Profilometer tests rung on HPD assembly 2E, Rod B, indicate that localized clad deformation occurs during operation. (2) Task 1B-Fuel Fabrication Development. Assembly. All fuel irradiation has been terminated with the final shutdown of the VBWR. The highest average burnup achieved by a single assembly in the group was assembly 4S with 8400 MWD/T. All assemblies in the group have failed or are suspected of failure. The Phase I developmental fuel continues to be irradiated in the Big rock Point reactor with the lead assembly having reached 1500 MWD/T. Fifteen phase II developmental assemblies are being construction for insertion at Big Rock Point in March. Engineering is underway to provide one instrumented assembly probe and two spare flowmeters for …