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We are enclosing three copies of a proposal concerned with the preparation for preliminary evaluation of up to 128 Hanford CV size I and II fuel elements to determine an optimum gas-pressure-bonding cycle for the fabrication of this type of fuel element. This proposal is based on your request as outlined in your letter dated May 7, 1962, and document No. HW-72417. Included is a cost estimate of the funding required to conduct this experimental study. All of the aluminum-clad uranium fuel assemblies are to be prepared and furnished by HAPO for gas-pressure bonding at Battelle in existing equipment. To implement the proposed program we have also enclosed six copies of the BMI-AEC research agreement (Serial No. 88). The agreement provides for a six-month research period with an estimated total cost of $24,780. This sum includes $1,180 find fee as established in Contract No. W-7405-ENG-92. Receipt of two fully amounted copies of the agreement, classification guidance, and the specimens will allow us to proceed.

ABS>The preparation of germanium hydrides, by the dropwise addition of al alkaline solution of hydroborate and germanate to aqueous acid, was studied systematically. As much as 70% of the germanium in solution could be converted to germane, Digermane, trigermane, and a polymeric germane were also obtained, and the infrared absorption spectra of gaseous trigermane and of polymeric germane were recorded. (auth)

The rear face or discharge area of a reactor contains all the appurtenances necessary to discharge irradiated fuel, to collect hot coolant from each process tube, to monitor tube and effluent temperatures, and to monitor the coolant for ruptured fuel elements. Generally, failure of a rear face piping component would not affect the safety of the reactor since the coolant has fulfilled its purpose, that of cooling the fuel elements. The failure may, however, cause failure of one of the monitoring devices and if undetected could lead to a minor reactor incident. The Purpose of this report is to review all information generated during the past three years concerning the condition of rear face piping and hardware. This review includes the history of rear face piping and hardware problems, study activities taken to ascertain the condition of the components, action taken to correct actual component failures, programs recommended to correct deficiencies which operating experience and engineering judgement indicate are necessary, and programs to accumulate additional information to support design of new piping and hardware components.

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The results of Spert I in-pile transient tests of a rodtype, low- enrichment UO/sub 2/ fuel element are presented. The tests were performed to investigate the possibility of damage to such long thermal-time-constant fuel rods when subjected to short-period power excursions, and to test the effectiveness of an instrumentation technique for measurement of UO/sub 2/ fuel temperatures within the rods. In an initial series of power excursion tests, in which the range of reactor periods was from approximately 1 sec to 7.5 msec, simultaneous measurements were made of the transient temperature at the center of the fuel rod and at the outer cladding surface. Fuel rod rupture occurred during the exponential rise of the 7.5-msec excursion. Similar short-period tests performed on a second fuel rod contain ing no internal thermocouples did not result in cladding failure, supporting the postulation that rupture of the first rod was caused by waterlogging of the UO/sub 2/ as a result of the cladding penetrations made for installation of the internal thermocouples. Calculations of the transient temperature distribution in the fuel rod were made, and the results are found to be in good agreement with the experimental data obtained on the central-UO/sub 2/ and cladding-surface …

This test is identical to the original except that it authorizes the performance of a trial reduction in reactor flow during a prior reactor shutdown. This trial flow reduction will be performed in the same manner as proposed for the actual test, with one exception. This is, that based upon the results of this preliminary test some changes in the timing of the different steps may be indicated. Such changes can readily be handled by making each step dependent upon the observed reactor outlet temperature during the test performance. The other significant change in the production test is the increase in the allowable bulk outlet temperature from Ti + 40 {plus_minus} 3{degrees}C{sup *}. This change is needed to obtain a reasonable extrapolation of the results of tests No. 1 and No.2 to 90{degrees}C, and is justified from a hazards standpoint by the excellent flow control achieved during test No. 1 and by the trial test that will be run prior to the performance of the actual test No. 2. Other aspects of the test basis and justification are presented in the original production test.

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gone approximately 16 years of almost continuous irradiation. Throughout this time stored energy has accumulated at a slow rate to the present maximum value of about 35 cal/gm releasable to 250 deg C. A small portion of the moderator (approximately 4%) contains stored energy which under adiabatic conditions may be released spontaneously (at approximately 165 deg C) to produce a maximum temperature of 270 deg C. Careful analysis has shown that the presert condition is not hazardous; however, it appears wise at this time to initiate some corrective action (thermal annealing) to prevent the continued buildup of stored energy to a dangerously high value. Several methode of obtaining effective annealing in the OGR were investigated. The proposed method was selected upon the basis of convenience, over-all safety, effectiveness, and cost. The proposed method involves the alteration of the present coolant flow system to permit reversal of air flow through the fuel channels. This will result in a reversed temperature distribution wherein the maximum graphite temperature will occur in the normally cold, maximum-stored-energy region of the moderator, Such an arrangement permits an annealing operation to be performed under conditions very similar to those of the normal safe operation. The proposed procedure …

A brazed end closure is currently being used on prototypical NPR fuel elements. The production closure will use a braze alloy composed of 5% Be + 95% Zry-2 to braze the Zircaloy-2 cap to the jacket and to the metallic uranium core. A similar MTR test, a GEH-4-57, 58, used a braze alloy of the composition 4% Be + 12% Fe + 84% Zry-2 which melts at a lower temperature. In this previous test, element GEH-4-57 failed through a cladding defect located at the base of the braze heat affected zone. Because of this failure it would be desirable to subject a fuel element, which had been subjected to more severe brazing conditions, to the same conditions as GEH-4-57, 58. For this reason the thermal conditions of this test essentially match those of GEH-4-57, 58. This irradiation test consists of two identical fuel elements. The fuel material is normal metallic uranium, Zircaloy-2 clad of the tubular geometry, NPR inner size. The fuel was coextruded at Hanford by General Electric`s Fuels Preparation Department. Each element is 10.8 inches in length with flat Zircaloy-2 end caps brazed to the jacket and uranium core with the 5 Be + 95 Zry-2 brazing alloy, …

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