This report is about the United States Self-Judging reservation to the compulsory jurisdiction of the international court of justice.

Program 3 of Operation BREN consisted in a study of the performance of biological shields composed of various arrangements of steel, water, and borated polyethylene in spherical, cubical, and parallelepipedal geometries. Studies were also made of the radiation levels within concrete-lined and -capped pits, 4 ft in dia and 8 ft deep. The principal radiation source was the ORNL Health Physics Research Reactor, an unshielded, unreflected, 90 wt % U-10 wt % Mo, cylindrical bare-metal assembly. It was supported at heights ranging from 27 to 1500 ft by a hoist car riding outside of a 1527-ft-high steel tower. A 1200-curie Co/sup 60/ gamma-ray source, similarly supported, was substituted for the reactor for some measurements. The work reported includes measurements of fast-neutron and gamma-ray dose rates and thermal-neutron fluxes in air and within the various shields, using the reactor source, measurements of gamma-ray dose rates in air from the Co/sup 60/ gamma-ray source, and measurements of gamma-ray pulse-height spectra within the pits, also employing the Co/sup 60/ source. 8 refs., 33 figs., 29 tabs.

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The release of uncontrollable quantities of harmful fission products such as I{sub 131} upon processing areas. At present this is done by procedural techniques using cards, filing techniques, and cross checks. These methods are entirely independent of the characteristics of the fuel itself and are subject to failure from human error. It is therefore advantageous to find a method of determining the age of irradiated fuel after discharge from the properties of the fuel. Such a method, with its relative immunity to human error, would be very useful in checking shipments as well as for such jobs as determining the correct age of fuel being stored for accumulation of desirable fission product daughters. It was the purpose of this effort to make a study of possible methods of direct fuel age determination, formulate and perform a test of the most promising method, and analyze the results in terms of practicability here at Hanford.

During the major part of October the N-Reactor was shut down for dump condenser repair, scheduled maintenance and equipment modifications. Work on dump condensers was successfully completed and the construction and tie-in of steam generator cell No. 4 was accomplished. Power ascension was initiated on October 24 and the N-Reactor in now operating with all five steam generator cells in full service. Process physics, in-plant testing, and technical activities are described in this document.

This monthly report details activities of the N-Reactor Research and Engineering Department during the month of December 1964.

The N-Reactor was shutdown on July 7, 1964, due to fuel rupture indications on tube 3361. The rupture has been confirmed as a failure in the downstream inner fuel element in tube 3361 caused by a pin hole in the weld and braze metal. This document presents a historical record of the rupture indications observed prior to and following the shutdown. The operating history of tube 3361 is compared with an adjacent tube, with no abnormal conditions observed at any time. Reactor temperature and power history since initial primary loop heatup are presented to show the total operating history experienced by N-Reactor.

This paper contains the minutes of a meeting on Columbia River contamination at Richland, Washington, held on July 16, 1964. The US Public Health Service sponsored the meeting and representatives from the following organizations were present: Washington State Pollution Control, Washington State Department of Health, Richlands Operations Office, AEC Headquarters, and General Electric Company. The meeting dealt with reducing the amount of effluent contributed to the Columbia River, the amount of reduction achievable, the costs involved and the status of technical feasibility studies. This document was declassified on January 26, 1994.

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The C-1 Loop is a small in-reactor water loop located in the 105-C Building. The loop was originally designed and constructed for nonfueled testing of the effects of reactor radiation on sample corrosion, crud deposition, and coolant radiolysis. The loop was modified in 1964 to permit the irradiation of aluminum-clad fuel elements and subsequently used for about eight months for the in-reactor testing of aluminum-clad, plutonium-aluminum fuel assemblies at high coolant temperatures and pressures. The facility is currently inactive but is available for the performance of special testing. Prior to nuclear operation of the C-1 Loop the mechanical equipment was carefully checked and the backup coolant system was experimentally verified to perform as designed. A safety analysis was also made with results summarized briefly in Reference and in more detail in the production test which authorized nuclear operation of the C-1 Facility. This report updates those earlier documents with improved descriptions of the C-1 Loop and its operation including analyses of postulated accidents.

Abstract: This report presents steam cycle optimization studies for large sodium graphite nuclear power generating stations.

BS>Testing of type 304-L austenitic stainless steel was completed. Thermal-stress-fatigue testing was started on ferritic pressure-vessel steel (ASTM type A302, Grade B), and work on this alloy is approximately half completed. Test specimens of type 403 martensitic stainless steel were machined and the heat treating schedule worked out in andticipation of testing this steel. The machining of test specimens of Croloy 2 1/4 steel was also started. (auth)

A series of irradiation tests were performed on the High Flux Beam Reactor fuel elements. The U--Al fuel elements contained 30 to 40 wt% U and 3% Si. Burnup ranged from 20 to 50%. A summary of the results is given. (C.E.S.)

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