Abstract: A stainless steel clad 9-rod assembly fabricated by The Babcock & Wilcox Company was irradiated in a boiling water loop of the General Electric Test Reactor. A post-irradiation examination revealed no significant dimensional changes on the fuel rods. the results of mass spectrometric analysis made of the pelletized UO2 fuel indicated a maximum burnup of 11,500 MWD/tonne was attained by Rod B-4 during the exposure.An x-ray diffraction examination of an unirradiated fuel sample revealed the presence of UN2 and U2N3 phases. Metallographic examination of the irradiated microstructures revealed similar second-phase particles.

Technical report describing that as part of the AEC Fuel Cycle Program, tests are being conducted to evaluate the significance of current fuel design limitations that do not permit the maximum fuel temperature to exceed the melting point of UO2. The reliability of prediction of the fuel rod operating conditions that will cause melting of the UO2 was evaluated by means of a calibration test conducted in the Vallecitos Boiling Water Reactor. Conclusions: (a) The central portion of the 3.15-cm diameter uranium dioxide fuel column melted. It appears that the UO2 was molten to a radius of 1.22 cm in the peak power region. The maximum extent of melting probably occurred during the peak power run when the kdT in this region of the rod reached 171 watts cm. The estimated radius of melting from metallographic examination indicates the kdT for sintered UO2 is 89 watts/cm. This supports a calculated estimate for sintered UO2 thermal conductivity published by D. R. deHalas and G. R. Horn. The results of the previous calibration run and subsequent experimental data by Lyons are also consistent with the value. This conclusion is contingent on the interpretation of the post-irradiation crystal structure of the UO2. Insufficient …

Abstract: A core design study of a 10 Mwe fast ceramic reactor is presented. Local reactivity coefficients, safety criteria, accident analyses, and economics are considered. An attempt is made to find a new balance of characteristics by purely geometric devices, i.e., by exploring the sodium : fuel ratio and varying the height : diameter ratio of the core. The use of BeO in the core was also investigated.

From introduction: "This is the seventeenth in a series of quarterly reports which cover the progress and results from the conceptual designs, economic evaluations and research and development work performed by the General Electric Company as part of Contract AT(01-3)-189, Project Agreement No. 13."

The influence of the Doppler effect in the core disassembly process following a meltdown accident is examined with a Bethe-Tait type model in which the Doppler effect, as well as core disassembly, is considered in the reactor shutdown process. It is shown that a strong negative Doppler effect can radically reduce the explosive energy release in such an accident. (auth)

Abstract: Critical experiment studies were performed, varying the parameters U235, B10 and metal to water ratio, in the SM-2 7 x 7 core configuration with 38 stationary elements and seven control rods of the SM-1 (APPR-1) type. An experimental mock-up of the SM-1 was assembled using the basic SM-2 fuel plates. Excellent agreement between the SM-1 boron loading, determined by chemical analysis, and the SM-1 mock-up boron loading, for equivalent bank positions, was noted. Several SM-2 mock-ups, cold clean and midlife, were assembled and studied with regard to reflector effects, flow divider effects, relative control rod array worths, critical rod configurations, and relative power distributions. The results of these experiments indicate as satisfactory a U235 loading of 36.4 Kg and a B10 loading of 63.4 grams for the SM-2. Attention is drawn to numerous power peaks present in the active core. The open seven control rod array has a slight reactivity advantage over the closed seven array and consequent minor disadvantage with respect to "stuck rod" criteria.

The hydrogen contents of several types of fabricated uranium have been determined by a vacuum method and expressed in terms of ccH2/ccU. The data indicate that alpha-rolled metal contains about 0.25 ccH2(STP)/ccU whereas beta heat-treated uranium yielded values between 0.30 and 0.37 cc per cc. Restricted efforts were made to determine where in the heat treatment the 5 to 10 cc of hydrogen per slug were taken up. It appears that no one operation is wholly responsible for this additional gas, although reactions between beta heat treated surfaces containing microscopic defects, and nitric acid may possibly play a large role. In general it may be said that slug produced by powder metallurgical techniques contain less hydrogen than pieces produced by rolling and heat treatment.

Abstract: The results of activity buildup studies in the SM-1 performed during Core I lifetime (June 3, 1957 to April 28, 1960) are reported. Data are presented on the extent, nature, and mechanism of the buildup of long-lived gamma emitting nuclides in the reactor primary system. Radiation levels after reactor shutdown are presented, as well as mathematical equations used to account for the observed activity levels. The data have shown that Co60 is the major contributor to radiation levels in the SM-1. Co60 activity arises from the cobalt in Haynes 25 alloy flux suppressors, and the cobalt impurity in stainless steel. After 35 months operation at an average power level of 55%, deposited Co60 activity accounted for approximately 83% of the total radiation level (mr/hr) contributed by the long-lived gamma emitting nuclides. The contribution of the primary coolant activity to the total radiation level is insignificant when compared to the contribution of the activity deposited on the walls of the system. The radiation level on the super-heater side of the steam generator was about 1400 mr/hr after 35 months of reactor operation. The percentages of Co60 activity in the coolant and in the deposits were not the same. This indicates …