ARMY REACTORS PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1962

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; 8 7 < 8 < : : : 6 9 9 = < 9 < : 5 < > ;" icipation in the program continued to include review, inspection, and support in various areas of reactor technology. An advanced fuel irradiation test program was established that is to be conducted in the pressurized-water loop in the Oak Ridge Research Reactor. Review of the design of the MH-1A reactor was initiated. This reactor, a pressurized-water system fueled with low- enrichment bulk UO/sub 2/ clad with stainless steel, is being designed as a floating plant to furnish electrical energy to shore ... continued below

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Pages: 127

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Creator: Unknown. April 10, 1963.

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; 8 7 < 8 < : : : 6 9 9 = < 9 < : 5 < > ;" icipation in the program continued to include review, inspection, and support in various areas of reactor technology. An advanced fuel irradiation test program was established that is to be conducted in the pressurized-water loop in the Oak Ridge Research Reactor. Review of the design of the MH-1A reactor was initiated. This reactor, a pressurized-water system fueled with low- enrichment bulk UO/sub 2/ clad with stainless steel, is being designed as a floating plant to furnish electrical energy to shore installations. Studies of the out-of-pile corrosion resistance of stainless steel brazed joints were completed. T-joint specimens of type 304 stainless steel were brazed together with 18 different alloys. Initial testing resulted in the selection of five of these alloys for extended testing, which was carried out in autoclaves with O/sub 2/ or H/sub 2/-O/sub 2/ added to the autoclave water. These alloys, General Electric alloys Nos. 81 and 75, Coast Metals alloy NP, low-melting Nicrobraz, and a Pdbase alloy, were satisfactory. Coast Metals alloy NP was selected as the reference braze material for the SM-1 fuel elements because it was more amenable to the brazing method established for this element. A method was developed for stabilizing Eu/sub 2/O/sub 3/ against aqueous corrosion in the event a control rod containing this material became defective in service. The method developed involves the addition of MoO/sub 3/ to the Eu/sub 2/O/sub 3/. When a mixture of Eu/sub 2/O/sub 3/- 15 wt% MoO/sub 3/ is reacted, a single phase inert to aqueous corrosion, believed to be Eu/sub 2/MoO/sub 12/, is formed. The irradiation testing that was carried out on nonsintered bulk UO/sub 2/ fuel in the ORR pressurized-water loop was completed, and the loop is to be used for irradiation testing of advanced cermet fuel elements, as well as water chemistry experiments in support of the Army pressurized-water reactor program. The design of the cermet fuel assemblies is under way. It is expected that experimental assemblies will be placed in the loop during the spring of 1963. In the meantime, the water chemistry studies now under way will be continued. The ORR loop has operated satisfactorily since December 1959 in the ORR. This loop was designed to recirculate water at conditions up to 2500 psig, 650 deg F, and 80 gpm. A recent test to determine the loop heat exchanger capacity at a nominal operating temperature of 480 deg F indicates that it is capable of removing 300 kw. The operating experience of the loop has been very good. Water leakage is extremely low, and thus excellent control of the water chemistry is obtained. While much information on water chemistry is already available, relatively little in the way of a fundamental understanding of observed behavior has been gained, especially in connection with the important unsolved problems of the transport and deposition of corrosion products and their associated radioactivity. At the beginning of the water chemistry study in the ORR pressurized-water loop in early 1961, cooled water samples taken from the loop upstream of the ion exchangers showed relatively low activity levels during normal undisturbed loop operation compared with similar samples taken during loop startup. About 90% of the activity was removed by the purification system ion exchangers, and about two- thirds could be removed by a 0.45 mu millipore filter with a cellulose backing pad. The crud level was estimated to have been < 0.05 ppm during normal operation. In contrast, during loop startup, the water-borne activity level was about 50 times higher and the crud level was often several ppm; 98 to 99% of the activity was removed by the ion exchangers, and 95% could be removed by a 0.45- mu filter. After the middle of 1961, radiochemical analysis of the water samples taken upstream of the ion exchangers showed a decrease in the level of water-borne activity. These later samples generally showed a lower percentage of activity removed by a 0.45- mu filter. It appears that the normal crud level in the ORR loop since the

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Pages: 127

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  • Other Information: Orig. Receipt Date: 31-DEC-63

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  • Report No.: ORNL-3386
  • Grant Number: W-7405-ENG-26
  • DOI: 10.2172/4704598 | External Link
  • Office of Scientific & Technical Information Report Number: 4704598
  • Archival Resource Key: ark:/67531/metadc1034312

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  • April 10, 1963

Added to The UNT Digital Library

  • Oct. 18, 2017, 7:39 a.m.

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  • Oct. 24, 2017, 3:41 p.m.

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ARMY REACTORS PROGRAM ANNUAL PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1962, report, April 10, 1963; Oak Ridge, Tennessee. (digital.library.unt.edu/ark:/67531/metadc1034312/: accessed November 18, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.