Discussion of laboratory tests to reduce uranium corrosion.

Fourth part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).

Cycle study performed on a closed loop Brayton cycle incorporating a gas-cooled nuclear reactor, regeneration, and compressor intercooling. The purpose of the study was to determine the effect of various cycle parameters on efficiency and cost.

Abstract: An analysis of SM-1 Core II and SM-1A Core I zero power experiments was made by comparing these cores to each other and to AM-1 Core I on the basis of critical bank positions, bank calibrations and available chemical analyses of the fuel plate compositions. The effects of replacing boron absorbers by europium absorbers upon rod worth and stuck rod conditions were studied. Comparisons of measured and calculated power distributions were made. It was concluded that both SM-1 Core II and SM-1A Core I contain nearly identical B-10 loading of 17.79 grams, compared to the best estimate of 15.75 grams for SM-1 Core I. The available data indicates that all three cores possess similar nuclear characteristics.

Abstract: The removal of both SM-1 Core I high burnup elements from SM-1 Core II and the insertion of the PM-1-M-2 element and the SM-1 Core I spare element in SM-1 Core II is discussed. Nuclear and thermal characteristics of Core II with these changes are presented and conclusions related to the changes in the hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself leads to a more conservative estimate of core safety. There is no conceivable reason why the perturbation may not be safely made in the SM-1 Core II.

Abstract: This technical report elucidates principles of hazards evaluation. The concept of hazards potential is introduced and utilized to show how a reactor system perturbation will influence its nuclear safety. Literature relating to reactor safety is referenced to provide the sources of information required for hazards analysis and show how they influence a hazards evaluation. A checklist of items which should be considered in evaluating a change, test, or modification is presented.

Phase I - Plant Transient Analysis. Behavior of the basic and refined kinetic models differs only slightly. It is therefore suggested that the basic model be used in any studies where the improvement in fidelity attainable fro the refined model is not warranted by the complexities introduced by the addition of function generator to the analog circuitry and derivation of the function to be programmed. The parameter responses of both kinetic models appear to be essentially similar to those of the plant with the exception of the primary pressure. In the pressurizer analysis it was noted that the primary system pressure surges of the model should be higher than those of the plant because of the adiabatic steam compression assumed in the model derivation. the fact that the mode surge is very much greater indicates that the compression process is far from adiabatic. A more detailed and complex model of the pressurizer, one that evaluates heat transfer in the steam pocket boundaries, would therefore reduce the costly conservatism otherwise necessary in the specification of pressurizer vessel size. Phase II - Xenon Reactivity Transients. On the basis of this study an analog computer circuit has been presented which accurately represents bank …

Report documenting radioactive fallout across the world, including material in the atmosphere, on land, and that has deposited into food and water supplies.

Report discussing the results of experiments on an EOCR prototype control rod, control rod drive, and driver fuel assembly that were extensively tested in 500 deg F Santowax and 132 deg F water. These tests not only established the operating characteristics of these assemblies but also revealed a number of deficiencies in the various components. It includes descriptions of the control rod, control rod drive, and driver fuel assembly together with descriptions of their method of operation.

Second volume of reports on the operation of nuclear test reactors. It includes six chapters: engineering description fo the materials testing reactor, reactor control components and instrumentation, reactor control circuitry, reactor operation, reactor shutdown and tank work, and supplemental facilities at the materials testing reactor.

Report describing the status of the SM-1, SM-1A, and PM-2A reactors, specifically regarding the effects "of irradiation on nil-ductility transition temperature and the associated problem of brittle fracture." (p. iii)

Report from Hanford Laboratories concerning "the design and construction of the 100-N Reactor and heat dissipation plant complete with the necessary auxiliaries" (p. 2). Details of its construction and the plant's systems and instrumentation are described as well as economic considerations.

A simple, light-weight, split beam optical smoke densitometer is described.

Thesis discussing experiments to separate azeotropes into pure components testing eleven minimum boiling point azeotropic systems: (1) acetone-methanol, (2) carbon tetrachloride-ethyl acetate, (3) cyclohexane-ethanol, (4) benzene-isopropanol, (5) benzene-n-propanol, (6) benzene-cyclohexane, (7) benzene-methanol, (8) benzene-ethanol, (9) acetone-carbon, (10) isopropoanol-water, and (11) ethanol-water.

From introduction: "The purpose of this report is to describe the observations made during the post-irradiation examination of HPD-2S, and to discuss possible modes of failure.

From preface: "The purpose of this report is to present an engineering test program for investigating a loss of coolant accident for a pressurized water-cooled and moderated reactor system."

Equilibrium data are presented for the distribution of thorium nitrate and nitric acid between aqueous solutions and tri-n-butyl phosphate diluted with kerosene.

Report documenting research and developments made by the Health Physics Division of the Oak Ridge National Laboratory.

Report containing supplementary material for "Plutonium Recycle Critical Facility, Final Safeguards Analyses." This material expands upon previously supplied safety information and resolves previously unreviewed safety questions. Appendix begins on page A.1.

Critical review of the information on zirconium, with estimates of the uncertainty in the values.

Data on the heat generated in irradiated uranium represented in both curves and formulas.

Derivation and basic formulas the physicist or nuclear engineer needs to do the necessary nuclear calculation for a reactor design.

From introduction: "Complete exploitation of the advantages of thorium as a fertile material in thermal and particular resonant spectrum reactors has been inhibited by the lack of quantitative physical property data for thoria and thoria-urania fuels."

Liquid Metal Fuel Reactor Experiment program quarterly technical report.