This report describes the operation of the fuels preparation department for the month of August, 1957. Manufacturing, employee relations, process development, plant improvements, and financial operations are described.

The purpose of this report is to present a basis for the application of slug corrosion rate data to corrosion limitations on pile power levels.

This document is the progress report for the design and development of Pile Area {open_quotes}G{close_quotes}, to cover the period June 1, 1949 to December 31, 1949. This project represents the major effort of the Reactor Division of the Design and Construction Divisions. It is being pursued with the aim of being able to incorporate the major project objectives in a finished pile design if that design is begun about January, 1951.

We have prepared what we call a bonded segmented fuel element. This fuel element has characteristics which make it an attractive prospect for high specific power operation at relatively low uranium temperatures. The fuel slug is made up of six longitudinal segments; each segment being bonded to the aluminum webbing and tubing that encases it. The pertinent information for this irradiation is summarized along with fabrication details and hazards, thermal aspects and calculations, and figures.

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Eight Zircaloy-2 jacketed, natural uranium seven-rod cluster elements were irradiated in a KER loop to determine flow channel temperature characteristics. One of the elements, which had 200 MWD/T exposure, was sent to the Radiometallurgy Laboratory for examination in April 1959. An outside rod of the cluster was sectioned and examined metallographically. No cracks or flaws were observed in the uranium cladding or bonding.

The purpose of this supplement is to authorize adjustment of the fuel loading to maintain a balanced longitudinal flux distribution in the region loaded for this Production Test.

Defect testing of the irradiated specimens will provide kinetic autoclave defect test data on the behavior and performance of coextruded Zr-2 clad fuel failures; this should reveal the effects of exposure on the coextruded U-Zr-2 bond.

The objectives of this supplement are to test an improved prototypic slug column displacement winch and associated remote control equipment and to demonstrate the feasibility of exerting control over the front-to-rear neutron flux changes as an aid in controlling reactivity cycling. An array of poison displacement columns up to twelve in number is intended, each charge consisting of a center section of I and E natural uranium slugs with mint pieces on each end. Initially six such columns will be charged with the remaining six being added later if required for proper control. The initial testing of the slug column displacement winch has established that the mechanics of the system are practicable and has indicated that the system would be feasible. The next phase of this test is to demonstrate the feasibility of controlling the front-to-rear flux changes over the entire reactor and to verify the improvement in equipment design. To be effective in controlling front-to-rear flux changes, there should be poison displacement tubes available in each quadrant of the reactor. It is believed that a maximum of twelve such tubes will be sufficient to show effective control with six of these being charged initially and the others added if …

The purpose of this test was (1) to determine the conversion ratio of alternated enriched uranium fuel segments and lithium aluminum target slugs, and (2) to determine the stability of solid and cored enriched uranium with this type of load. The test consisted of an irradiation of two twelve-tube pile charges, each of them forming a sixteen tube square without corners. The first twelve tubes were a ``striped`` load containing solid fuel elements enriched to .95% U-235 alternated with lithium-aluminum target slugs. The fuel elements of the second array were 1/2 inch cored, but canned with the core plugged to give the same O.D. and external appearance as a solid element. To determine fuel element stability, the irradiation was essentially a run-to-rupture test. The central four-tubes of each square were designated for special extraction after the irradiation to determine conversion ratio. Neither solid nor cored enriched uranium of metal quality comparable to that used in this test has adequate stability to reach a routine goal exposure of more than 500 MWD/T at the specific powers of the test, 70--80 KW/foot. From post-irradiation measurements of diameter growth of the cored slugs, it might be concluded that many of the slugs were …

The possibility of overboring in existing Hanford reactors to increase slug size has been suggested as a means of increasing production. The effect on control system strength of such a modification is, of course, one of the factors which must be considered. This report presents the results of a study requested by the Reactor Physics Unit of the local buckling of control rods as a function of the migration area. The migration area varies for different lattice make-ups, but due to the complexity, calculation of T and L{sup 2} for changes in slug and hole size was not done here; instead, the range of values given by the ``Giant`` Program was assumed.

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The status of the overall schedule for design, fabrication, and installation of the component items of the PAR loop experiment in the ORR has been brought up to date and reviewed. This is the first review since the original issue of the schedule

Objective of this production test is to obtain irradiation experience wit the hot-headed closure for co-extruded Zircaloy-2 jacketed rod. Seven Zircaloy-2 jacketed natural uranium seven-rod cluster elements with hot-headed end closures and spark machined end supports will be irradiated in the KER Loops to an exposure of 2000 MWD/T in pH 8--11 coolant.

This report presents preliminary engineering analyses of costs and possible capabilities for recovering gross and mixed fission products at Hanford. Three alternates were considered for recovering gross or mixed FP from Purex 1WW in B Plant: gross FP recovery as an individual total program, gross FP recovery as an incremental part of a waste solidification program, and recovery of a mixture of zirconium-95 and niobium-95 as an individual total program.

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1951. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

It is recommended that the following described I & E fuel element pair be evaluated for rupture rate improvement, and if significantly improved, recommended for C Reactor usage. Operating conditions are estimated.

The purpose of this report is to outline the preliminary design and provide justification for the installation of crossunder lines at B, D, DR, F, and H Reactors.

A metallurgical bond can be effected between a Zircaloy-2 jacket and a thin walled uranium tube by treatment in a high temperature, high pressure gas autoclave. Bonding is independent of core history, provided that all surfaces are free from contamination. Gas pressure bonding causes the fuel element jacket to conform to the core, which promotes bonding on all surfaces including the ends. Grain growth occurred in some instances between the jacket and end cap interface. A three phase system (alpha, delta and epsilon) was identified across the interface at the bonding conditions of 845 C (1500 F), 10,000 psi and a time of four hours. Four structures were observed; these were epsilon, epsilon plus delta, delta, and alpha plus delta. Microhardness tests were used to identify the various phases. Also, thickness measurements were made of each component in the interface. The normal pressure bonded interface consists mainly of epsilon and delta phases. A cast bonded sample, subjected to the pressure bonding treatment, consists predominantly of alpha plus delta material while an interface bonded during casting is primarily alpha plus delta phases and epsilon phase.

Several safety infractions at the plant are reported. Budget data is given. Some planning information is given for Purex design testing, separations process and equipment demonstration, separations chemistry, separations engineering, separations hazards, hydriside development, waste handling, and recycle development.

The objective of this supplement is to authorize a change in the panellit trip range from 25--75 psi to 5--95 psi. The test hole facility consists of two concentric aluminum tubes which extend from the front face to the rear face of the reactor. The ID of the inner tube is 2--7/8 inch. Water from one crossheader supplies the annulus, water from another crossheader supplies the inner tube. The three-foot-long, .570 inch OD fuel element is centered in a 40-inch long aluminum holder which has an ID of 1.380 inch and an OD of 2.800 inch. The panellit gage which monitors the flow to the inner tube fluctuates to such an extent during start-up that on two occasions the reactor was scrammed. During equilibrium operation the panellit gage reading remains stable. A possible explanation of this behavior is that during start-up aluminum spacers which are in the inner tube as part of the test charge chatter and cause variations in the water path through the tube. It is further surmised that at equilibrium operation the pressure drop across the column in the tube is sufficient to suppress the chattering. It is concluded that extending the trip range to 5--95 psi …

This report discusses the service life expectancy of the 105-N Reactor, Steam generation transients following a scram, and estimated number of outages.

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A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.