This document details activities of the Fuels Preparation Department during the month of December 1958. (FI)

To evaluate the self-supported fuel element concept, tests are underway to determine the performance of collapsible bridge-rail supported fuel elements in ribless process tubes under present reactor conditions at B Reactor. It appears expedient, however, to extend this evaluation to future operating conditions in order to establish the relative feasibility of conversion to self-supported fuel elements in ribless tubes in present reactors. Utilization of 1.47% U-235 enrichment will provide fuel element powers comparable to those attainable under proposed future conditions. Since I & E fuel elements of this enrichment have previously attained exposures in excess of 2000 MWD/T at specific powers averaging 75 KW/ft in C Reactor, this test will specifically evaluate the feasibility of the self-supported fuel element concept. The purpose of this report is to present the design of a test to fabricate and evaluate self-supported fuel elements under conditions of comparable severity to those expected for future loadings of this geometry.

A solid spline made of 430 boron stainless steel was tested in tube 2451 at D Reactor. The test was conducted to compare the boron carbide filled hollow aluminum spline presently used with a solid steel spline which is being considered as a substitute. Usage of hollow splines has shown that a solid rugged spline would be highly desirable. Comparisons of spline radiation and reactivity effect were obtained both in-pile and in out-of-pile test facilities. Results of the production test are reported in this document.

This report discusses testing three-foot Zircaloy-2 jacketed tube-and-tube elements with either natural or 1.6% enriched unalloyed or 2% zirconium alloyed uranium which will be irradiated in any of the KER Loops to goals up to 5000 WD/T. Downstream thermocouples to measure flow channel coolant temperatures will be included for charges of enriched elements. Special tray discharges to prevent the fall of the elements to the basin are required for all loadings. The tube-and-tube geometry is being investigated for use in the NPR because of its large surface to volume ratio and low uranium temperature for a given specific power. A series of irradiations is planned to study the dimensional stability and uranium behavior of the tubular elements at jacket surface temperatures up to 315{degree}C and exposures up to 5000 WD/T. Because of its lower reaction rate with high temperature vater, the uranium-2 w/o zirconium alloy may be useful in a high temperature recirculating system. Consequently the behavior of this alloy will be compared to the behavior of unalloyed uranium.

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Last year (1948) during the flood period the flow at Grand Coulee fluctuated widely. 2 PM, June 8, 543000 c.f.s.; 4 AM, June 9, 568000 c.f s.; 2 PM, June 9, 543000 c.f.s.; 2 AM, June 10, 573000 c.f.s. A total instantaneous fluctuations of 37,500 c.f.s. was reported. Now there is installed a new control. This control can keep downstream variation within 500 c.f.s. By lowering the lake level prior to the crest period, the drum gates could be used as flood control (1948 high water basis) the drum gate control plus the water turbine discharge (if the lake level had been reduced) could have dropped the crest at Richland three feet. a. Drop in crest at Richland one foot: Electrical loss nominal, b. Drop in crest at Richland two feet: Electrical loss 1 megawatt/foot for six generators. Loss Max possible 13,310 KW each generator, 79,860 KW total (7 days). Capacity 1,170,000 KW Max Loss 6.8% for 7 days to 10 days. c. Drop in crest at Richland three feet: Electrical loss 1 megawatt/foot for 6 generators Max possible 30,100 KW each generator 180,600 KW total 8 days. Capacity 1,170,000 KW Maximum loss 15.4% for 8 to 12 days. Actual …

This document details activities of the Fuels Preparation Department during the month of November 1959. (FI)

Sections of the dummies furnished by Radiological Engineering, Process Reactor Development Operation were machined into 1/4 inch diameter by 1 inch long cylinders and irradiated in the Quickie Facility at F area. The pieces were discharged directly into a holder one foot from the Beckman chamber. The transient time from in-pile to the chamber is approximately 30 seconds. The readings were taken using a Beckman chamber, Beckman Micro-Micro Ammeter and Recorder. This system has been calibrated with Co{sup 60} sources obtained from the Oak Ridge National Laboratory. We are including data taken from a sample of 61-S and 99.998 per cent aluminum which may be of interest.

It is the objective of this study to investigate preliminary expansion program requirements for process water, as supplied by 190 Building equipment; from the point of view of practical pumping, flywheel and pump suction head requirements. These requirements are to be determined at this time in such a form and accuracy as to be useful in refined estimating for budget study purposes. In order to obtain the objectives of this study at this time it has been decided to consider five different conditions of process water flow to a reactor. These conditions are flow to the reactor under summer conditions of operation in gallons per minute with a corresponding top of riser pressure in pounds per square inch: 85,000 gal/min, 580 psi; 100,000 gal/min, 480 psi; 130,000 gal/min, 280 psi; 150,000 gal/min, 280 psi; and 150,000 gal/min, 150 psi.

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Prior to this experiment, no measurements of Ball 3X effectiveness had been performed for any of the older 2004-tube Hanford piles, and calculations concerning total control requirements were made assuming the vertical safety system strength equal to the strength of the B, D, F vertical safety rods. With current and projected enrichment loadings, the vertical control system was calculated to be inadequate to satisfy the total control criteria at all times, resulting in the necessity to provide supplementary control in the form of horizontal rods or temporary process tube poison. Because of the larger ball channels at DR, the Ball 3X system is stronger than at the other 29-VSR piles. Therefore, the potential relaxation of total control limits was greater for DR should an experiment show the Ball 3X strength to be significantly greater than the B, D, F VSR strength. PT-IP-126-C authorized an experiment to determine the Ball 3X effectiveness at DR Pile by measurement of the relative strength of a VSR and a column of 3X balls in the same channel. The experiment was performed in March, 1958, and on the strength of favorable results a supplement to the PT was prepared to authorize the same test at …

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 objectives of this test are to obtain a current knowledge of the H Reactor flow system characteristics under various transient conditions which have a reasonable probability of occurrence. The test data will be used to revaluate: (1) the H Reactor bulk effluent temperature limits, (2) the margin between the Ball 3X riser pressure trip settings, and (3) the quantitative adequacy of emergency flow available from the combined high tanks export system. This test will consist of two parts: (a) Measurement of the flow characteristics during transition from the primary coolant system to the secondary coolant system (190 steam driven pumps.) (b) Measurement of the adequacy of the last-ditch emergency water system (combined high tanks-export system) during a simulated primary and secondary coolant system failure.

This monthly document details activities of the Hanford Operations Office during the month of October 1959. (FI)

This test is designed to accomplish two primary objectives: (1) to attempt to verify ex-reactor corrosion data which indicated improved corrosion resistance of cast blank X-8001 alloy material compared with wrought blank, and (2) to attempt to verify the resistance to groove pitting type of corrosion attack previously observed on M-388 components. Ex-reactor tests of cast blank M-388 alloy in autoclaves using water as the corrosive media up to 360 C, and in flow loops up to 120 C have indicated that the corrosion resistance of the cast blank material is equivalent to, and probably superior to the corrosion resistance of wrought blank material. Metallographic examination of this material indicated a more uniform nickel dispersion in the aluminum as a probable explanation of this performance. The sporadic occurrence of severe groove pitting has seriously challenged the use of X-8001 nickel aluminum alloy as a fuel element cladding material. Although the actual cause of the groove pitting has not been determined, non-uniform dispersion of the nickel in the alloy is suspected. The cause of the non-uniform nickel dispersion or segregation has been located and virtually eliminated by removal of additional aluminum (scalping) from the ingots prior to fabrication of the components.

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This document details activities of the irradiation processing department during the month of January 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering operation; Employee Relations Operation; and Financial Operation.

This document details activities of the Hanford Operations Office during the month of February 1959. (FI)

The objective of this production test is to determine the changes that occur in the uranium-Al-Si bond during irradiation at bond temperatures between 255 and 285 C. Twenty-five M-388 jacketed dip canned depleted uranium solid fuel elements will be irradiated to an exposure of 500 MWD/T in high temperature water. The location and size of unbonded areas on the fuel elements will be measured by ultrasonic mapping before and after irradiation to show the changes in bonding resulting from irradiation at high temperature.

Two coextruded, Zr-2 clad, natural uranium, seven rod cluster fuel elements were irradiated to a calculated exposure of 1250 MWD/T in the KER Facility and discharged 1-16-59. The fuel elements were NPR candidate fuel and examination was requested to determine the behavior of coextruded, Zr-2 clad, natural uranium irradiated at core temperatures of approximtely 425{degree}C. The elements were transferred to the Radiometallurgy Laboratory 2-25-59. The elements demonstrated excellent in reactor performance with no significant changes in either the fuel or the hardware. Detailed examination of the central rod and two peripheral rods from one of the clusters showed no microcracks in the uranium. Moderate growth was observed in the fuel at the unrestricted rod ends.

The purpose of this report is to present the results of experimental tests made to determine the degree of protection offered by the Panellit system to accidental flow reductions caused by plugging incidents to a K reactor process tube operating at low front header pressures.

The objective of this Supplement described in this report to Pt-IP-250-A is to d enriched tube-and-tube elements will develop pitting corrosion on the Zircaloy-2 jackets when irradiated in pH 10 water. The measurement of dimensional changes in the fuel elements and the observation of the effect of irradiation on the uranium and bond area are also objectives of the test, but secondary in importance to identifying a pitting corrosion problem in NPR quality water, if one exists.

The purpose of this report is to provide scope and justification for the installation of two new 16,500 gpm, 150-foot head pumps in the 181-C building in order to furnish sufficient additional water to increase production at 105-B and to supply the entire normal export demand from B Area. A minimum of 11,000 gpm will be required to be transferred from the 183-C reservoir to the 183-B clearwell via the cross-tie line during the adverse turbidity period when the I&E program is in effect at 105-B. It is contemplated to transfer 15,000 to 20,000 gpm so as to provide the necessary water for this increased flow at 105-B and to reduce the flow through the existing 183-B filters. It is apparent that the pumping capacity at 181-C is not adequate to supply this extra water at 183-C unless all export pumps are used to pump water to 183-C. If this increased demand were met by using the export pumps, the 100-D Area would have to supply most of the export demand to the 200 Areas which is anticipated to be 12,000 to 15,500 gpm. After the new expected I&E flows are achieved at 105-D and 105-DR, D Area will have …

This document details activities of the Fuels Preparation Department during the month of January 1959. (FI)