A manual is given of the equipment and procedures used in the Army Reactor (SM-1A) to control the water purity and makeup. In addition to a description of the primary purification control system, a discussion is presented of the water chemistry control procedures for the auxiliary systems (e.g., the spent-fuel pit, the shield tank, and the waste disposal system). (T.F.H.)

An investigation was conducted to establish an operating procedure for flushing water through a purification system demineralizer which was out of service for an extended period, and to determine the demineralizer serviceability. (J.R.D.)

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The feasibility was studied and a cost estimate prepared of an experimental reactor to determine the operating characteristics of beryllia- moderated, gas-cooled systems wtthin a power limit of 10 Mw(t). The heat energy produced by the experimental reactor is to be dissipated in a heat dump. No machinery for production of power was to be provided. Other requirements were that the reactor should be capable of testing core types different from the current MGCR design, and the system should permit use of gases other than helium. It was further directed that the reactor should be designated BORE for Beryllium- Oxide Reactor Experiment. Reactor development work was mainly in connection with the BORE preliminary design. It was established that the most important information which could be provided by a 10 Mw(t) reactor experiment would be on performance of fuel elements and moderator bodies. This required that the experirment duplicate the power density in the fuel and moderator that would exist in the full size reactor and made it advisable to use full length fuel elements. This resulted in an unconventionally shaped core which is roughly cylindrical with the length more than twice its mean diameter. Studies continued on performance of fuel …

Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high order of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to …

This document details activities of the Hanford Laboratories Operation for the month of August 1960.

This report presents a proposed test program to study the corrosion of aluminum-clad fuel elements under conditions expected at the more severe conditions which will be encountered in operating the present Hanford reactor at higher power levels.

The possibility was reviewed that errors in calculations might have resulted in an erroneously high theoretical strontium activity which would explain the unexpected low activity found ion the strontium recovered from Purex wastes. The new, corrected calculations (49,213 curies/T) improves the accounting from 65 to 75% of theoretical. The discrepancy still should be investigated.

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This document details the activities of the Hanford Operations Office during the month of September 1960. (FI)

This report discusses activities from the fuels processing department. Activities described include personnel statistics concerning injuries and accidents; operating plans concerning the N loading activities, 305 test reactor, projection fuel elements; process fuel element testing, and general engineering operations from the department.

This report addresses the critical mass studies, part X.

Reactor maintenance problems and engineering studies have identified various reactor rear face piping problems at 105-B, D, DR, F, and H. The long-range solutions to these problems are outlined in classified document HW-65269, ``Existing Reactor Rear Face Piping Review.`` These problems, however, must also be considered from the short-range view to minimize their effects on continuity of reactor operation during the interim until the permanent remedy can be accomplished. This is expected to be a minimum period of approximately four years. Flow and bulk outlet temperature up to 90,000 gpm and 95{degrees}C respectively are assumed to prevail in these short-range considerations. The purpose of this report is to review Reactor, Plant Engineering`s efforts on the short-range aspects of this problem.

This report, for August 1960 from the Chemical Processing Department at HAPO, discusses the following: Production operation: Purex and Redox operation; finished products operation; maintenance; Financial operation; facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

Proposed fuel element loading and operating conditions for this test are tabulated. The loading is designed to provide high tube power and near-maximum-allowable coolant outlet temperature, accompanied by low fuel surface heat fluxes. Conditions which would exist during the proposed test were compared with those during the rapid depressurization incident in KER-2 on Jan. 31, 1960.

This document details activities of the irradiation processing department during the month of August, 1960. 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.

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.

This monthly document details activities of the Hanford Operations Office during the month of August 1960. (FI)

A resume of Health Physics activities for April, May, and June, 1960 is presented. Discussions and tabulations which summarize results of field surveys, bioassay, personnel monitoring, and environmental surveys are included.

Ability of the originally designed SM-2 stainless steel fuel elements to withstand static transverse pressure gradients of normal reactor operetion was investigated. Fourteen simulated original SM-2 type fuel elements by Tungsten inert gas (TIG) and electric resistance welding processes were subjected to pressure gradients to 6 psi and measured for deflection during stress and distortion after stress. Tests provided a basis for selecting a reference SM-2 fuel-element design and established a method of predicting maximum plate deflections under anticipated SM-2 operating conditions. Results showed that TIG welded elements, slightly more prone to deflection, are preffered for SM-2 application because they are less susceptible to weld failure and that fuel plate thickness is the most significant parameter in determining plate strength. It was also found that equilizing the length of the outer and inner plates significantly reduces maximum plate deflection and that effects of weld spacing and side plate thickness were not significant within the pressure range of the tests. (auth)

Computational results for the characteristic values and solutions of Bessel's equation are presented. Specifically, the results pertain to the first two solutions subject to the Neumann boundary condition at r = a,b and to the first solution subject to the Dirichlet condition, when (b-a)/(b+a) is small (0.1, 0.01, and 0.001). Approximate analytic formulas are derived to complement the computational results. The possible application of the data to phenomena involving the interaction of an intense circulating beam and the electromagnetic fields within a particle accelerator is indicated, as is also the utility of the results for estimation of the loss-factor, Q, for resonant electromagnetic modes of possibly high order within a toroidal vacuum chamber of rectangular cross section. (auth)

Reactor design information obtained from 3 years' operation of a small pressurized-water reactor, the SM-1 (formerly APPR-l), is presented and discussed. The SM-1 reactor, designed to produce 10 Mw(t) power, employs fully enriched uranium fuel in the form of UO/sub 2/ dispersed in stainless-steel fuel plates. The reactor is cooled by water at 1200 psia and mean temperature of 44) deg F. Core-physics measurements were performed of temperature coefficient, pressure coefficient, rod calibration, stuck rod position, and transient xenon as a function of core burn-out. Core burn-out characteristics were compared with few- group calculations, and reasonable agreement was obtained. Thermal-heat-balance data were obtained on the reactor core. The temperature pattern in the nominal and hot channels under operating conditions was calculated. These calculations indicated that certain of the fuel channels operated in the nucleate boiling regime. Examination of one of the fuel channels suspected of nucleate boiling indicated no adverse effects. The system response to load perturbations and during pump coast-down was measured utilizing plant instrumentation. This response was compared with analytical predictions using a lumped kinetic model, and reasonable agreement was found. Both neutron and gamma traverses were made through the primary shield during reactor operation. Gamma traverses were …

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