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This report addresses a brief review of he international symposium held in Albuquerque, New Mexico.

The oxidation of graphite in the Hanford reactors is of consequence since graphite burnout affects the strength of the moderator. As a means for indication of any highly oxidizing condition within the stack, containers or boats of small weighed samples of reactor-grade graphite are positioned along the length of an empty process channel in each reactor. The rate of oxidation of the monitoring samples, referred to as the burnout rate, is reported as percent weight lose per 1000 operating days (%/KOD). Currently the burnout rate limit is 2%/KOD. This document presents recent burnout data at the C-reactor.

This report is about the Classical Dynamics of Triatomic Systems which is divided into three parts.

The preliminary design is described of equipment to carry out experiments on boiling burnout in which the average coolant density as a function of coolant channel length as well as local coolant densities may be measured. It appears that by use of the equipment, average densities can be accomplished in a few seconds, while determination of a complete density map (with 5% error) across a plane of the assembly requires approximately one hour. (J.R.D.)

This report addresses the control design review of P&W quarterly for July to September 1960.

From time to time, in discussion of possible naval-based nuclear ramjet missile systems, a question arises concerning whether or not a particular reactor would or would fall within the `Tory II-C Technology`. Such a question is meaningless unless a definition is furnished. This report provides such a definition.

Fuel element warp occurring during the irradiation period is considered to be one of the major fuel element dimensional stability problems. Warp has been shown to correlate with accelerated corrosion attack. and also can contribute to stuck fuel charges, particularly in bumpered or self-supported charges where the annular clearances are reduced due to the presence of the projection rails. Thus, any process which offers a potential for reducing the average warp should be evaluated. Preliminary tests offsite have indicated that the use of a commercially available oil for a quench medium following beta heat treatment produces a fuel core with less residual stresses and a slightly finer and more uniform grain size than that produced by the present HAPO method of water quench. Thus, adoption of an oil quench nay offer a means whereby warp can be reduced without incurring costly revisions to equipment or fabrication processes. This report presents an irradiation testing program to evaluate the performance of oil quenched cores and to determine the optimum core heat treatment.

The nuclear heat generation rate in the first capsule irradiated was higher by a factor of two than was calculated. The original capsule was irradiated in a dry bore with cooling water in the annulus only. The new capsule will be irradiated in a water-cooled bore facility with additional cooling coils around the lithium containing tube. This will keep the inner capsule temperature below 150 C during the initial tritium buildup period prior to outgassing. This Supplement authorizes the irradiation of an additional capsule and the removal of the present facility and installation of a single tube general purpose facility. All remaining provisions of the original development test are in force except for minor exceptions due to the water flow in the bore which have been changed in the following writeup.

Currently the B, C, KE and KW reactors are in operation. The B and C reactors are operating to produce 12% Pu-240 in both the E metal (0.95) and natural loadings. Some columns will be taken to 15% Pu 240 on a test basis. Production of 12% Pu-240 will result in exposures of 1400 MWD/ton on natural metal and 2,000 MWD/Ton on E metal. Currently exposures are such that 10 1/2% Pu-240 has been achieved. Some control problems and temperature cycling have been encountered with this mode of operation. Both K reactors are on enriched-depleted loads. The depleted metal is scheduled for irradiation to produce 27-30% Pu-240 (about 3,000 MWD/Ton; 15 months residence) and is currently at 30% of goal. The E metal drivers (0.95) will be discharged at weapons grade Pu exposures. The relatively small quantity of natural uranium in the K reactors will be exposed to produce 12% Pu-240.

The specific impulse achieved in a pulsed rocket engine augmented with a fissioning nuclear bomb could be greater than that of any continuous flow engine. To a certain extent, this increase in specific impulse would be obtained at the expense of motor weight and average thrust. This paper considers the first of these limitations, motor weight, and estimates the highest effective specific impulse to be expected from a nuclear-pulsed rocket motor with respect to the weight of the motor.

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Production Test IP-546-A. consisting of 20 charges for the purpose of irradiating the first hot die sized fuel elements was charged in C-Reactor July 27, 1963. A second tests PT-IP-616-A containing a total of 19 charges had been charged on two separate outages; 13 land 6 tubes on September 28, 1963 and October 30, 1963, respectively. Fabrication of test quantities was becoming more routine and additional irradiation tests were planned. This report summarizes the fabrication of hot die sized fuel elements originally intended for Production Test IP-630-A, but changed to ``Supplement ``A`` to Production Test IP-546-A, Irradiation of Diffusion Bonded Fuel Elements,`` HW-75465 E. The production test was changed due to unexpected growth behavior of hot die sized fuel during irradiation.

The Heating and Ventilation System provides the atmospheric conditions in FFTF as required to assure that plant objectives can be met in a safe and practical manner. Included are functions and design requirements, a physical description of the system, safety considerations, principles of operation, and maintenance principles.

The FFTF Radioactive Waste System provides the equipment and facilities required for the routine collection, transfer, storage, and disposal of radioactive liquid and solid wastes. Radioactive solid wastes and radioactive liquid wastes not normally piped t o the Radioactive Waste System Receiving tanks, are packaged inappropriate containers by the systems generating the waste for collection and disposal by the Radioactive Waste System. Included are functions and design requirements, a physical description of the system, safety considerations, principles of operation, and maintenance principles.

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A program for the irradiation of samples-of N-Reactor graphite has been in progress since November 30 1961, in the General Electric Test Reactor (GETR). The basic purpose of the program is to achieve an exposure to the samples equivalent to 20 years of operation of N-Reactor and to relate the physical distortion of the samples to the expected behavior of the N-Reactor moderator stack. Theoretical studies to relate sample exposure in the GETR to exposure in N-Reactor have been underway since the program began. The conversion of exposure depends on three main items: the flux and exposure in the GETR test position; the flux and exposure in the N-Reactor stack; and the relationship of the amount of carbon damage per unit flux in N-Reactor to that in the GETR. To date all three items of the conversion have been based strictly on computer calculations. The first item (the GETR flux) has been calculated as group one, E > 0.18 MeV, of the three-group computation made every reactor cycle by the GETR physics group, This computation is normalized to reactor conditions by comparison of group three, E < 0.17 ev, with measured cobalt-alloy monitors. No direct comparison of group one with …

A commercial graphite useful in nuclear reactor construction is described. A survey of all currently available sources on chemical and physical properties was made and the information listed. Some data on cost and available sizes are also included. (auth)

A half reactor test was initiated at B Reactor on February 1, 1967, to determine whether the concentration of P{sup 32} and As{sup 76} in the reactor effluent water would be reduced by the addition of activated alumina in the water treatment process. This test was prompted by the reported effectiveness of activated alumina in sorption of phosphates when water flow is through a packed bed of this material. Although direct feed of material would not be expected to be as effective, testing of this possible approach appeared warranted. This report summarizes the results of this test. Samples taken from each effluent riser during reactor operation indicated no improvement in either As{sup 76} or P{sup 32} performance. This comparison is shown below with a complete tabulation of daily values included as Attachment 1.

Metallic thorium uranium fuel elements continue to show excellent irradiation performance in high temperature pressurized water coolant. Volume expansion measurements made after 2.1{times}10{sup 20} fissions/cm{sup 3} (6200 MWD/T) at fuel temperatures above 500 C shows no indication of fission gas-in-duced swelling. Analysis of fuel swelling data from tubular elements shows apparent effect of geometry (restraint) on both low temperature volume expansion and the temperature at which accelerated volume expansion initiates. Evidence of grain boundary tearing has been observed in tubular metallic uranium fuels irradiated in 1600 pos water coolant. Volume expansion due to two types of structural damage have ben observed uranium - 2 w/o zirconium alloy irradiated to 0.25 a/o burnup at temperatures up to 550 C.

Purpose of this supplement is to increase the maximum allowable liquid level in the near downcomer for single downcomer operation. The required clearance between the downcomer lid and liquid level is reduced from 24 inches to 18 inches in the near downcomer, providing bulk outlet temperature is maintained less than 93 C.

The described information consists of fuel element specifications, irradiation system parameters, operating conditions, and failure observations to include rupture identification.

With recent concern for downcomer capacity at most areas, it seems appropriate that a survey be made to determine what instrumentation and data are available. The following is a brief description of the available instrumentation and data for KE, KW, H, F, DR, D, C, and B reactors.