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This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic·Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during August 1969 .

Abstract: RFD-2 was the second of Sandia's operational safety flight tests of systems for nuclear auxiliary power.

Nine inch long aluminum clad Li-Al alloy target elements contained in 35 miI wall Zr-2 cans are authorized for irradiation in KER-1 and KER- 2 for up to 150 equilibrium operating days. The Li-Al elements will be irradiated in 1.9 inch OD, 1.5 inch ID Zr-2 flow distributing sleeves. Normal N inner fuel elements in Zr-2 sleeves will included in the tube loadings to provide sufficient heat to maintain the loop at the desired operating temperature.

This report is divided into: present reactor fuel production, NRD production fuels, N fuel development, and current reactor fuel development.

The following data represents processing conditions used in fabricating prototypic (except for supports) natural and enriched NPR element assemblies and K-type I&E{sup 2} for the following production tests: ETR, 6{times}6 Loop; GEH-10, Nos. 52, 53, 54 and 55; KER Loops; PT-377; MTR Loop; GEH-4, Nos. 68, 69 and 70. The purpose in documenting the data is to provide a permanent record of processing conditions and dimensions which may be referred to for post irradiation analysis and possible future process work. Post irradiation results will be issued by the Fuels Development Operation, Hanford Laboratories Operation, and the test loop operating conditions will be issued by Process and Reactor Development Operation, Irradiation Processing Department, as outlined in the Production Test Procedure.

The control rod calibrations in the E-N core loading and in the natural uranium core with an E-N blanket yielded results which indicate that the controls are approximately 10 per cent less effective in an E-N core due to shadowing of the rod strength by the E-N lattice; it had been expected from theoretical, calculations that the E-N shadowing effect would be approximately 10 per cent. The tests thus indicate that there has been approximately 25 per effective rod calibration value since the 1957 calibration due to changes in the flux pattern associated with alterations in fuel type and geometry. The significant difference between the measurement made in 1957 when the loading was basically natural uranium, and the measurement made in November 196l, when the pile contained E-N fuel, is thus concluded to be a combination of the blacker lattice and flux distribution effects.

The purpose of this report is to describe the results of prototypic tests of piping alterations designed to alleviate the above described conditions and to outline a program for full-scale implementation of these alterations.

The purpose of this document is to provide a tabulation of data concerning ruptured slugs for the use of various interested parties. This report presents data for the ruptures that occurred from January 1, 1962, through April 30, 1962, inclusive. Definitions of the terms used in the tabulations are nearly self-explanatory and are documented in the reference document. The possible exceptions to this are the three terms: type of metal, type of failure, and lot number. The legends currently in effect for these three terms are provided in this report.

It was stated by HAPO representatives at meetings held in Washington and Savannah River on May 1st and 2nd, that roughly 200 kgs of Pu of 18 percent Pu-240 could be furnished by July 1, 1965 to Savannah River for their proposed Cm-244 program. The Pu would be produced by irradiating .2 percent U-235 depleted uranium for a period of roughly ten months. The reactivity required for the depleted uranium irradiation would be furnished by .947 percent U-235 enriched uranium. Nine tons of slightly enriched uranium would be required for each ton of depleted uranium during its complete irradiation. To implement this program to deliver Pu of high isotope content by July 1, 1965 will require immediate internal program approvals and appropriate implementation steps. These steps will include design and specification of the depleted uranium ordering appropriate quantities of E and depleted; determining and ordering appropriate canning components; and determining reactor loading patterns. Special attention will ultimately be required in the separation of the reactor products. The additional quantities of E-metal required for the loading will probably necessitate early installation of the annular type Purex dissolver.

Tube-tube type fuel elements require support devices to maintain concentricity both between the element components and between element and process tube. A variety of support devices have been conceived. all require welding of a tab or other attachment device to the Zircaloy clad. Ultrasonic and resistance welding were evaluated for this application. Investigation showed ultrasonic welding to be not acceptable in its present state. Resistance welding proved satisfactory. This paper will briefly cover the work done in evaluating the two methods of attachment.

Nickel plating on the coolant contacting surfaces of aluminum-jacketed fuel elements is highly attractive for increasing resistance. Potential benefits include a highly corrosion-resistant coating for severe localized conditions, reduction of mechanical damage to fuel element jackets, improved fuel element alignment (by reducing friction between fuel element and process tube ribs) and probably lower overfall surface temperatures to reduction in corrosion product film with improved corrosion resistance, neutron economy might also be realized. For example, substitution of a 0.5 mil thick nickel plate for 15-mils thickness of aluminum jacket would result in no reactivity loss and permit a concomitant increase in uranium volume, or in coolant flow annulus. Attendant problems include providing an adherent continuous plate of uniform thickness and possibly contamination of reactor effluent by radio-nickel-cobalt, and phosphorous and it was found that gross sloughing of the nickel plate had occurred. Development and testing work was carried out to determine the cause and a solution to the Greece problem. Studies were limited to the behavior of chemically-deposited nickel because of the unique capability of the process to deposit a coating of uniform thickness in the 0.1 - 0.2 mils thick range, regardless of the geometry of the plated piece. Based …

This report discusses a comparative irradiation test of fuel elements of N-fuel uranium composition at Hanford.

The Purex L-Cell Package is used for organic stripping and concentrating aqueous plutonium solutions from the final Pu solvent extraction decontamination cycle. At the present time the package is used only during times when the gamma activity of the final product is too high to permit routing the solution to N Cell for final ion-exchange purification or when N Cell is inoperative. The package can be used only to concentrate and then route solutions to the PR Room for loadout into PR cans for storage prior to rework via TK-E6. In addition, operating data have shown that the present package has a low boil-off capacity which is insufficient for both current and forecasted plant production rates. At the present time, flowsheet modifications and processing rate reductions are necessary when the package is in operation. Entry into L Cell and contact maintenance are required to maintain the present package which was installed in 1956. Repairs to the package were made in 1962, and information obtained at that time indicates that extensive repairs to the present package will be required in the near future. The purpose of this report is to present the engineering study for the required modifications to the package …

The test parts were canned on October 7, 1962 and October 8, 1963. Only six parts, all radiographic rejects, were used for the test instead of 12 parts as originally planned. The original test called for using one very porous and one relatively non-porous sand cast aluminum insert in each container. Shipping experience has shown no significant effect of porosity of inserts on corrosion severity of parts shipped in them. All inserts used were full contour support inserts. Each test part was canned in a separate container under conditions described in the process test. The shipping assemblies, were stored for 21 days and the parts removed and inspected on October 29, 1963. Colored photographs of the parts were taken on October 30, 1963.

In reviewing the data on hydrogen content in the canning baths, the solubility limit of H{sub 2} in Al-Si, and the U-H{sub 2} relationship as was suggested earlier, it appears that the surface H{sub 2} on the core should be less than 0.4 ppm to be below the H{sub 2} concentration level routinely found in the canning baths. What effect low surface H{sub 2} of about 0.4 ppm would have on braze porosity after canning is unknown. However, several years ago (about 1955) some solid cores were vacuum outgassed at MAPO to about 0.5 ppm total H{sub 2} and very little braze porosity resulted.

In phase I tungsten clad cermet fuel specimens were thermal cycled, to study the effects of fuel loading, fuel particle size, stablized fuel, duplex coatings, and fabrication techniques on dimensional stability during thermal cycling. In phase II the best combination of the factors studies in phase I were combined in one specimen for evaluation.

This r e p o r t was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 f o r the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during November 1969.

Abstract: This report presents a summary and brief evaluation of the RFD-2 flight-test data reduced as of December 16, 1964.

Preparation of about 1200 g of uranium nitride powder is reported. Experiments indicated that synthesis of mononitride by decomposition of higher nitrides required pro longed heating at high temperature. Chemical analysis of uranium nitride for nitrogen was found to be difficult. Data from analysis of these compounds are tabulated for comparison with calculated composition. Sintering studies of uranium mononitride continued during the report period. Work on uranium silicide formation from the elements without quenching was continued, and effects of crucible material on the synthesis of the material were investigated. (For preceding period see ORC-254.) (J.R.D.)

This technical report is a summary of short time strength data for CS1830 series Cb-1Zr material. The data are presented as a function of fabrication history or the amount of cold work received during fabrication. Three fabrication categories were considered for comparison: forgings and extrusions; bar, plate, rod and pipe; and sheet and tubing. Forgings and extrusions having received no cold work showed the highest strength over the entire temperature range 68F to 2800F; bar, plate, rod and pipe received moderate cold work and showed lower strength than forgings and extrusions but higher strength than sheet and tubing in the temperature range 2000F to 2800F.

This Systems Study is submitted in accordance with the requirements of AEC Manual Chapter 1801, and proposes the procurement and implementation of a Plant Process Surveillance System for N Reactor. The envisioned system consists of instrumentation, data acquisition equipment, a digital computer, and input/output equipment. It is intended that the system will be a tool for reliable administration of safety limits and standards, simplification of plant operation, and reduction of operating costs by replacing the tasks of shift operations personnel, an instrument technician and a process engineer.

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