Introduction: The objective of Project Agreement 22 is to determine the feasibility of covering the complete reactor neutron flux start range from 10(3) - 5 x 10(13) nv by using in-core chambers. The counting mode of operating will be used at low neutron fluxes and the root mean square voltage fluctuation mode will be used at high neutron flux levels. Experiments have been run utilizing various ion chambers, gases, gas pressures, voltage, and cables to measure sensitivities and range operating in the counting and RMS voltage modes. Theoretical discussions are presented showing how the RMS voltage is related to individual pulse at both amplifier input and output. Noise is also compared at amplifier output so that the optimum bandwidth can be selected. Spectral shifts with changes in applied voltage causing signal variations have been examined and can be eliminated by appropriate selection of amplifier bandwidth. In the counting mode, all experiments have been conducted with unterminated cable. The chamber has been designed with geometry, gas, and pressure to completely stop fission fragments in the gas and hence maximize the charge generated in the chamber. Cables have been selected to minimize capacity. Various gases, pressures, and voltages have been used to …

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.

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.

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.

Abstract: Residual and fission gas release from UO2 were studied in the laboratory and in in-reactor experiments. Arc-fused powder and sintered pellets were used to determine the rate of evolution and types of residual gases as a function of temperature. Fission gas release was related to the average UO2 temperature and fission gas release calculations were made using the latest thermal conductivity values, isotopic half lives, and branching ratios available in the literature. The results obtained were compared with those available in the literature, and a satisfactory agreement was found among the groups of comparable data.

Summary: The Mixed Spectrum Superheater (MSSR) is an integral superheater reactor in which boiling occurs in an annular Boiling Water Reactor section and steam in superheated in an unmoderated fast section in the center. A Mixed Spectrum Critical Assembly (MSCA) to be operated at the Vallecitos Atomic Laboratory has been designed to mock up a 75-150 MWe prototype MSSR. The principal experimental measurements aimed at proving the feasibility of the MSSR concept include power distribution, Doppler effect, flooding effects, distribution of reactivity, control rod worths, and the effect of the control system on the power distribution.

Summary: Technical report describing the testing of eight Potter Meters, for metering inlet flow and measuring exit steam qualities in the Consumers Big Rock Point Instrumented Fuel Assemblies, were individually calibrated for flow and pressure drop up to 500 gpm in the low temperature (130 F) fluid flow facility. The flow calibration comparison made with an ASME orifice installation, agreed to within + - 1 percent among seven of the meters, and meter Serial No. 8 was 2.8 percent lower than the others. Pressure drop among the meters was within about 5 percent. Locked rotor pressure drop data was obtained on one meter. A fully instrumented fuel bundle was tested in the low temperature facility and pressure drop data obtained for the tieplates and meters, spacers, and channel rods. A mock-up of the exit end of the instrumented fuel bundles, composed of 1 foot of fuel rods, tieplate, and Potter Meter was tested in the High Pressure Heat Transfer Facility. Data was obtained for single- and two-phase calibration of total flow and exit steam quality in an instrumented bundle. Each meter was operated, for a minimum of 6-8 hours after bearing modifications necessitated by seizure of the rotors, in the …

From introduction: "This is the sixteenth 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."

Abstract: A stainless steel clad 9-rod assembly fabricated by The Babcock & Wilcox Company was irradiated in a boiling water loop of the General Electric Test Reactor. A post-irradiation examination revealed no significant dimensional changes on the fuel rods. the results of mass spectrometric analysis made of the pelletized UO2 fuel indicated a maximum burnup of 11,500 MWD/tonne was attained by Rod B-4 during the exposure.An x-ray diffraction examination of an unirradiated fuel sample revealed the presence of UN2 and U2N3 phases. Metallographic examination of the irradiated microstructures revealed similar second-phase particles.

Summary: The High Power Density Project includes a comprehensive fuel development program which has the objective of developing and demonstrating the performance of a nuclear reactor core having a high power density, long fuel life, and low fabrication cost. The fuel program is made up of two principal tasks. Task 1A consists of irradiation tests in the VBWR of Type 304 stainless steel clad, UO2 pellet type fuel rods fabricated by current commercial processes. Task 1B consists of the investigation of lower cost fabrication processes and the irradiation testing of fuel elements fabricated by these processes. Both tasks include the investigation of the feasibility and use of thin-wall stainless steel cladding as a means of improving the neutron economy and fuel cycle costs of stainless steel clad fuel. Irradiation of the Task 1A fuel assemblies in the VBWR was initiated in September, 1960. Subsequently, Task 1B fuel assemblies were inserted in the VBWR as various fabrication processes and design concepts were investigated. Fuel cladding failures have occurred in fuel rods in both Task 1A and 1B. As of this date, cladding failures have occurred in twenty-two rods of approximately 700 fuel rods which have been irradiated. Twenty of the failures …

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. The development of pulsed neutron source techniques for large power reactors has led to a new theoretical model recently developed by E. Garelis and J.L. Russell, Jr. The theory is presently based on a bare, one-group model with m-delayed precursors and takes all spatial modes into account. Results indicate, however, that the application of this model is much broader. Experiments were designed and carried out to both verify this new theory and to demonstrate the performance of the experimental hardware in a large power reactor.

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. The number of assemblies has been reduced to seven as a result of the failure of two pellet fuel assemblies. The average burnup of the group operating as of September 1 is 7500 MWD/T. (2) Task 1B-Fuel Fabrication Development. Assembly. Assembly 12S gave positive signals of being a leaker under the multi-type in-core sampler and was declared failed based on the in-core results and visual observation of a cracked rod. Modifications to the instrumented fuel assembly probes were made by removing the failed flow meter rotors to allow continued use of the flux detectors and thermocouples. Flux detectors and thermocouples performed properly after reactor start up. Flux wire tubes were found to be kinked such that their use was prohibited. (3) Task II-Stability, Heat Transfer and Fluid Flow. A series of noise recordings of fluxes, flows, and temperatures has been made at 91 MWt at the Big Rock Point plant. Preliminary analyses of some of the these records were made to obtain …

Technical report describing sixteen fuel rods clad with thin type 304 stainless steel and filled with vibratory compact powder UO2 that were fabricated and incorporated into a bundle for irradiation testing in the VBWR. The UO2 powders were tested for gas content. N2, CO, and H2 were the principal gases evolved by both type of UO2, but the arc-fused UO2 released about ten times as much gas as the Dyna Pak UO2. The amount of gas released was also a function of particle size and temperature. The gas evolution data were used to design the gas plenum to accommodate the absorbed gases along with the fission gases.

Technical report of a proposal of an analytical model to predict two-phase critical flow rate. The model is based upon thermal equilibrium, a "lumped" treatment of the two-phase velocity (each phase is represented by a single mean velocity), and upon the neglect of frictional and hydrostatic pressure losses. A comparison, of the proposed predictions with available test results and previous analyses shows that: (1) The present model agrees very well with the published test data. (2) In contrast to all other analyses, the model requires no assumption about the gas void fraction.

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.