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.

During the quarter the development program was continued as scheduled. Milestones that were reached included the successful irradiation of prototype Big Rock fuel bundles in VBWR to a level above 2500 MWD/T with power densities as high as 88 kw/1. Also completed was the scheduling computer specification followed by a formal solicitation of bids through the AEC. A summary of progress by task is presented here.

The Big Rock Point Reactor of the Consumers Power Company went critical for the first time on September 27, 1962. Summarized here are objectives and accomplishments of the pre-operational period as well as an indication of work scope remaining.

From introduction: "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."

The purpose of this report is to present a conceptual design study on a low power electrical and heat generating plant powered by nuclear energy. The nuclear reactor used in this study is the homogeneous circulating fuel type.

Experimental test loops have been used to study the problem of hydraulic stability. The oscillatory behavior of a single-phase and two-phase natural circulation loop has been examined at atmospheric pressures and oscillating modes of operation were studied in terms of heat input.

Introduction: The object of Project Agreement 22, Task 1, is to develop improved detectors which can operate up to 1000 F for in-core power monitoring. Several ideas have been developed to achieve this goal: (1) root mean square fluctuation voltage measurement of ion chamber signals, (2) thermocouple-type detectors, and (3) fabrication developments.

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.

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.

Abstract: This technical report covers the development work conducted during a planned program with the U.s. Atomic Energy Commission, Contract AT(04-3-189, Project Agreement 22, directed toward the development of high temperature, nuclear radiation resistant, telemetering devices. The development program is devoted to: (1) investigation and selection of two possible telemetering devices, and electromechanical commutating switch and an AM oscillator employing TIMM circuit elements, (2) procuring the electromechanical commutating switch to specification, (3) building and operating a TIMM oscillator, and (4) temperature testing of both devices. A resistance-coupled Wien-bridge sine wave TIMM oscillator was build and tested both as an oscillator, and in combination with other oscillators to simulate a telemetering system. An electromechanical commutating switch rated for 350 F operation, instead of 700 F as originally specified, was procured and tested. The drive motor and gear reduction unit which is designed to drive the commutating switch, is rated for 750 F operation and designed to operate in an nuclear reactor radiation environment of 1 x 10(17) nvt and 1 x 10(10) R.

The influence of the Doppler effect in the core disassembly process following a meltdown accident is examined with a Bethe-Tait type model in which the Doppler effect, as well as core disassembly, is considered in the reactor shutdown process. It is shown that a strong negative Doppler effect can radically reduce the explosive energy release in such an accident. (auth)

A reference design of a large high power density core has been established representing the available technology as of August, 1960. Reference core performance and cost should improve considerably after incorporation of improvements now under study.

Abstract. A program was instituted to study and reproduce the in-reactor intergranular failures of Type-304 stainless steel fuel cladding found in superheated steam. The program was directed toward finding ways to eliminate the cause of failure or to use improved alloys that would be less susceptible to failure. A materials screening test was developed in the out-of-pile superheat facilities with 1.5 ppm chloride added as sodium chloride to the recirculating water in the presence of typical boiling water reactor quantities of oxygen and hydrogen. During the test, the heater sheaths were exposed through several cycles to saturated steam (with its accompanying moisture carryover) and superheated steam. Failure of Type-304 stainless steel was obtained in periods of less than two weeks; the failures were predominantly transgranular. Type-347 and vacuum-melted Type-304 stainless steels failed in this NaCl-cycle test while Inconel-600, Incoloy-800, Hastelloy-X, Type-406 stainless steel, and vacuum-melted Type-310 stainless steel were acceptable. An improved chloride cycle test with 0.5 ppm chloride added as ferric chloride to the recirculating water was developed. An intergranular failure was obtained similar to that experienced in the superheat fuel cladding failures in the superheat in-pile loops in the Vallecitos Boiling-Water Reactor. Sensitized Type-304 and Type-316 stainless steels …

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.

Introduction: The purpose of this technical report is to review the water chemistry methods and equipment developed for use with the Maritime Loop Irradiation Program conducted in the General Electric Test Reactor (GETR) from December 2, 1960 to July 19, 1962. Special emphasis is given to areas having general application to other high purity water systems. The Appendix includes a discussion of specific conductivity and pH in high purity water systems. A major section of this report is devoted to a review of gross activity levels on coupons of two different surface finishes exposed in the loop coolant system for various time intervals. A major objective of the chemistry program was to select or develop analytical methods such that the analyses could be performed at the loop location by technical personnel who normally operate the loop. By this means, frequent samples were obtained and analyzed directly thus providing close monitoring and control of the loop water chemistry at minimum expense.

This report covers the S-I-5-B-M* fuel irradiation in the GETR Maritime Loop during the fourth quarter of fiscal year 1961. The data are summarized in Section II.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the final quarter of fiscal year 1962 plus the month of July 1962 to conclude the irradiation program. The data are summarized in Section II.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the third quarter of fiscal year 1962. The data are summarized in Section II. Discussions on fuel performance, fuel environment (water chemistry), and loop operations are included.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the first quarter of fiscal year 1962. The data are summarized in Section II.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the second quarter of fiscal year 1962. The data are summarized in Section II. Discussions on fuel performance, fuel environment (water chemistry), problems with loop operations, and the crud deposition program are included.

The General Electric Company is proceeding with an irradiation program to proof test a representative array of Savannah I fuel rods. Irradiation of a test assembly containing Savannah I fuel rods has begun and it is proposed that the results of this irradiation will permit an advance evaluation of the fuel performance and fuel burnup in the Savannah I reactor. This report covers the first two quarters of the reporting period. All aspects of the subject program have been consolidated and applicable portions are discussed in some detail.

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.

The experimental program described has had as its objective the determination of the pressure exerted by free fission gas in operating UO2-filled reactor fuel elements.