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 …

A research program is being conducted to obtain experimental data in the irradiation of plutonium-enriched fuel to confirm a theoretical model for predicting isotopic composition and reactivity changes in plutonium-enriched, light-water-moderated reactors. Quarterly progress: The densitometry procedure (for measurement of alpha autoradiographs of fuel pellets) has been modified to eliminate the need for a second emulsion. The existence of a problem of latent image fading and non-reciprocity of the high-resolution emulsion has been recognized. A tentative procedure has been worked out to correct these emulsion difficulties. the number of polished pellets has been increased to thirteen. The number of hot spots per pellet has not changed appreciably. The largest spot seen is irregular with an estimated volume equivalent to that of a sphere of 35 mil diameter with a PuO2 concentration in the neighborhood of 60%. The VBWR irradiation run now under way is not scheduled to end until October. To the end of the last run the cumulative exposure reached 3703 MWD/T, as logged by VBWR operating personnel. Applying the same scale factor between logged exposure and Ce-Cs analysis of the first fuel sample gives a corrected exposure of 4416 MWD/T. Further debugging of EPITHERMOS, the epithermal extension of …

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 that the pressure drop along an annular channel with dimensions D(1) = 0.375 inch; D(2) = 0.875 inch, L = 70 inches. Flow was vertical and upward, and only the internal surface was heated. Subcooled conditions existed at the inlet, with two-phase conditions at the exit. Groups of three radial spacer pins on 18-inch centers along the channel, held the inner surface concentric with the outer surface. The single phase loss coefficient for each spacer group is K(8) = 0.21. The single phase friction factor for the annual channel is given by f = 0.16 N(R)(-0.16). The two phase pressure drop increases as the quality increases for G [over] 10(6) = 0.5 ;b/hr ft(2). The effect of heat flux on the pressure drop is very is very slight over the range of fluxes tested (0.55 less than or equal to Q over 10(6).\ less than or equal to 0.8). The two-phase pressure drop gradient in the same annulus, with no heat addition is qualitatively the same as for a 1/4-inch by 1-3/4 inches rectangular channel but is quantitatively greater than for the rectangular channel.

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

Technical report describing the statistically designed empirical approach being used to choose a candidate Zr alloy optimum with respect to corrosion and hydriding rates in steam and having acceptable mechanical properties. The statistical techniques used and the reasons for their use are discussed in detail, with emphasis on the estimation of corrosion rates. Estimation of response surfaces is also considered.

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.

Abstract: (1) Grain growth in UO2 pellets was studies between 100 C and 2600 C. The pellets were encapsulated in small vacuum-tight tungsten containers in an argon atmosphere. (2) The grain size-time relationship could be expressed by an equation. A low exponent, m>_ 1/3, was found in those experiments and is related to the type of UO2 investigated. An activation energy of 65 kcal/mole was obtained for the grain growth process. The time exponent, m, increased with increasing temperature if the pellets were not contained in closed capsules bu heated under an argon pressure of 1.5 atm. (3) An interaction between tungsten and UO2 could be observed at a a temperature of 2600 C after prolonged heat treatment.

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: Two UO2-filled stainless steel clad fuel rods in which fission gas pressure was measured during irradiation have been subjected to post irradiation examination. Results of free gas analysis and metallographic examination are in general agreement with observed pressures reported previously. Calculated fuel surface temperatures based on extent of fuel recrystallization indicate that in a one-half inch diameter fuel rod with 0.014 inch diametral clearance operated at a maximum heat flux of 531,000 Btu/hr-ft, gap conductance increased with increasing heat flux. An analysis of void configuration indicates that pressure is more sensitive to as-fabricated void volume and changes in this volume resulting from fuel expansion than to fuel central temperature. The decreases in effective void volume with increasing fuel temperatures becomes more significant as initial void volume decreases, and excessive fission gas pressures may be developed in fuel rods operated at high fuel temperatures unless adequate expansion volume is provided in fabrication.

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 …

Technical report describing the four-rod heat flux experiments that are a part of a continuing program of study of the critical heat flux, or burnout phenomenon in order that water cooled reactors can be designed with a maximum of safety and efficiency. During heat transfer with boiling, there is a particular heat flux, for a given set of flow conditions and geometry, above which the nucleate boiling process begins to break down. This breakdown of the nucleate boiling process is known as burnout, critical heat flux, departure from nucleate boiling (DNB), and boiling crisis. The present method at General Electric of avoiding the critical heat flux conditions in the reactor is to limit the heat flux, for a given set of flow conditions, to a fraction of the critical heat flux at the same conditions in the single-rod test section of Janssen and Kervinen. Because the critical heat flux of a heater rod facing an unheated wall is lower than that of a heater rod facing another heater rod, the critical heat flux conditions of the single-rod test section, will be a conservative estimate of the critical heat flux conditions in a multirod reactor. The main purpose of these experiments …

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 …

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

Technical report describing that the pressure drops along 3/4-inch, 1-inch, and 1-1/4 inch straight pipes and across three contraction-expansion inserts in a 1-inch pipe have been measured under both single- and two-phase flow conditions. Pressure was varied from 600 to 1400 psia, flow from 0.25 x 10(6) to 1.66 x 10(6) lb/hr ft, and quality from zero to 90 percent. The single-phase pipe friction factor agrees with the Moody value for smooth pipe. The two-phase friction for horizontal flow shows no size effect in the range of pipe sizes from 3/4 inch to 1-1/4 inch. The values lie below the Martinelli curve at the lower qualities (x<0.6), but at high qualities tend to be above the Martinelli curve. The single-phase loss coefficient for the three contraction-expansion inserts show very little Reynolds number effect in the range of channel Reynolds numbers from 3 x 10(4) to 5 x 10(5). The two-phase data for insert number 1 has not yet been reduced. The two-phase loss for insert numbers 2 and 3 lies generally below the loss prediction based on a homogeneous flow model. The two-phase loss for insert number 2 shows excellent agreement with the corresponding loss for the S-1 insert in …

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

Task A: Modification and Preparation of Experimental Facility. With the exception of the insulation of modified components, the experimental facility is complete. Insulation will be completed by the end of September. the system has been charged with Refrigerant-22 and preliminary loop performance tests have been completed without the test sections.. Task B: Design and Construction of Test Sections. The stainless steel test section has been prepared complete with end flanges and pressure tap locations. Wall thickness tolerances have been ultrasonically checked. Test section inlet and discharge assemblies are being completed and the whole assembly will be ready for installation by the end of September. Glass sections from the same drawn length which will make up the final test sections have been received for pressure tests. The final coated sections and the associated inlet and discharge fittings will be ready for assembly by the end of September. The above sections were ordered after complete preliminary tests defined the properties required of these test sections. Task C: Design and Construction of Test Stand: The mechanical design and drafting of the structural components and drive system is complete. The electrical control system for the platform orientation has been constructed and two modes of …

Summary: Fundamental studies in support of the alloy design work are complete except for the experimental determination of the diffusion of oxygen in alloy-doped non-stoichiometric ZrO2. Over 100 oxidation runs have now been made on samples of ZrO2 doped with 1 mole percent of the oxides of Al, Y, Fe, Cr, and Ni. The first round testing of 31 alloys is now essentially complete. Analysis of the steam corrosion rate and hydriding raw data taken at 300, 400, and 500 degrees C indicates that Zr-Cr and Zr-Cu-Fe alloys show the most promise for development for service in steam over the entire temperature range 300-500 degrees C. Maximum resistance to corrosion hydrogen embrittlement requires high initial ductility and thus low, perhaps less than ~2.5 a/o total alloy content. For any composition, susceptibility to hydrogen embrittlement depends on crystallographic texture of the component; under certain circumstances hydrogen embrittlement may be high anisotropic. The second-round testing of 10 selected Zr-Cr and Zr-Cu base alloys is now about 50% complete. Three alternate fabrication schedules were evaluated; and the preliminary results indicate that the Zr-Cu alloy tested is less sensitive to heat treatment than is the Zr-Cr alloy tested. Raising the final alpha annealing temperature …

Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, fuel testing in TREAT, fuel performance evaluation, fast-flux irradiation of fuel, and reactor dynamics and design.

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

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.

The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.

Summary: Initial critical heat flux, transition boiling temperature fluctuation, and film boiling coefficient data have been obtained on a two-rod cluster assembly at 1000 psia and 25 to 90 percent steam qualities. A representation showing the range of critical heat flux data is presented. Typical temperature recordings which indicate transition and film boiling behavior are shown. Fabrication of a new high pressure observational test section is nearly complete. An optical table and illumination system has been build and operationally tested for photographic use on the new observational section.