A testing program designed to provide the necessary data for evaluation of the emergency flow backup adequacy was conducted at the K reactors on February 16, 1963. Three transient or trip-out tests and 13 steady-state tests were performed under authorization of PT IP-573-I. This report contains the numerical results of the pressure and flow data that were obtained during these tests.

This report explains a mathematical algorithm designed to complete contour descriptions of weapon components which have been only partially specified by free-form data.

In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

A test was performed at the Shippingport PWR, whose purpose was to determine the amount of heat the fuel shipping container cooling system was capable of removing while in a state of equilibrium. Studies were also conducted to determine if after four hours of operation, the increase of the internal wall temperature was less than 30 deg F, and finally, to determine the effects of loss of cooling water from the fuel shipping container. (auth)

A summary of studies conducted on the pressurizedwater reactor loss-of- coolant accident is presented, and an experimental safety program is proposed. The various phenomena involved in the loss-of-coolant accident, related research and development programs, assumptions currently used to predict the various physical phenomena, and the general approach to be used in conducting the safety tests are discussed. In order to accomplish the loss-of-coolant experimental safety program, a dry containment test facility is proposed for construction at the Test Area North of the National Reactor Testing Station in Idaho. The site selection utilizes existing support facilities suited for performing nuclear safety tests requiring experiment assembly areas and post-test analytical examination of the irradiated nuclear components. (auth)

Boron carbide-graphite nuclear control-rod materials containing up to 45 wt% B were prepared by hot-pressing techniques. The densities of these compacts decreased with increasing B content. Compact strengths, however, were noticeably increased with B content of 20 wt%, suggesting that the B acts as a sintering aid. Thermal stability was evaluated by annealing compacts under a variety of temperature and atmospheric conditions. The upper limit of the ioss of 13 was 5% to 10%~ in 200 hr at 2000 deg C--considerably less than would be predicted from the Langmuir evaporation equation. Diffusion of B atoms through the pores of the graphite, rather than evaporation from the surface, was found to be the rate- controlling step in the loss of B. Compacts containing 20, 30, and 40 wt% B were irradiated at 350 and 700 deg C to 1.0 and 2 4 x 10/sup 21/ nvt(fast). Physical integrity of the compacts was good after irradiation. Slight increases in tensile strength were noted in several instances. (auth)

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A program was written to perform polynomial least squares fits on the IBM 7090 computer. Weighting factors may be included with the data if desired. The program includes a subroutine which does the fit and furnishes the calling program with the coefficients, standard errors in the coefficients, staandard error of fit, and information necessary to compute a complete error analysis; and a calling program which reads input, calls the subroutine, and writes requested output. The subroutine may be used separately in any Fortran or FAP program where a least squares fit is needed. (auth)

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The radiation survey of the Blanket Exit Water Instrumentation assemblies Core I, Seed 1 was performed on January 15 and 16, 1960, approximately 100 days after reactor shutdown at 5806 EFPH. The observed radiation levels varied with height from 150 to 500 mr/hr; the highest activity was found at the bottom of both support tubes of each of the four assemblies monitored, which, in operation, would be located 5 in. above the top of the blanket fuel assembly. The two inner port Blanket Exit Water Instrumentation units and the two Auxiliary Seed Exit Water Instrumentation assemblies were not monitored. (D.L.C.)

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