This document presents the initial definition of a 2 Mwt, SNAP-50 Reactor Test and is intended and is intended to provide guidance to the Architect/Engineer in evaluating and costing various Nuclear Test Facility design concepts. The information is based on preliminary studies; and, consequently, the configuration and component sizes have not been firmly established.

This document presents the initial definitions of an 8 Mwt, SNAP-50 Reactor Test and is intended to provide guidance to the Architect/Engineering in evaluating and costing various Nuclear Test Facility design concepts. The information is based on preliminary studies; and, consequently, the configuration and component sizes have not been firmly established.

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This report details measurements of fission product ratios which were made an various samples of the Purex canyon exhaust filter to determine the manner and time interval in which the filter became plugged vith foreign material. Results for samples from different levels within the filter bed were compared to investigate accumulation as a function of depth. Strontium-89/strontium-90, Zr-95/Ca-137, Ru-103/Ru-106, and Ce-141/Ce. Pr-144, ratios were determined. These were compared to a constant rate model for accumulation of fission products on the filter. Observed Sr-89/Sr-90 and Zr-95/Cs-137 ratios fall within a range of values typical of normal plant operation, Specifically long-term accumulation of Hanford material irradiated at 10 MW/T to between 600 and 900 MWD/T total exposure and cooled between 120 and 200 days. Low Ru-103/Ru-106 and Ce-141/Ce, Pr-144 ratios definitely rule out rapid accumulation of short-cooled material as the build-up mechanism but are not sensitive indicators in the case of long-cooled material. Ratios did not vary significantly with sample location, indicating that accumulation age does not differ with depth within the filter.

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The objective of the test was to evaluate the behavior during irradiation of nominally 1.739 inch OD, 1.074 inch ID, Zircaloy-2 jacketed single tube fuel elements with brazed end closures at operating conditions somewhat more severe than those expected for N Reactor fuel element outer tubes. The test elements were fabricated with the Be-Zr alloy brazed end closure planned for use on N Reactor fuel elements. Iron supports were used to minimize process tube scratching during charge-discharge.

The feasibility of operation with an enriched core to ``override`` the xenon transient and thus reduce minimum downtime has been reasonably well established. The principle economic disadvantage of the present technique is associated with the use of boron splines for compensating the reactivity effect of the core during normal operation. The boron-containing spline parasitically absorb the excess neutrons produced in the core thus wasting these surplus neutrons during the operation when the enriched core is compensated by poison; also, because of the relatively slow removal rate of irradiated splines, a significant portion of the potential downtime reduction is used to remove the splines from the core. The purpose of this report is to point out incentives and alternate methods for improved poison control with the xenon override core and to present current plans for resolving these limitations to the xenon override concept.

The design of the High-Flux Isotope Reactor (HFIR) was supported by a series of preliminary experiments performed at the Oak Ridge Critical Experiments Facility in 1960. The experiments yielded results describing directly some of the expected performance characteristics of the reactor and strengthened the calculational methods used in its design. The critical assembly, like the reactor, was of a flux-trap type in which a central 6-in.-dia column of H/sub 2/O was surrounded by an annulus of fissile material and, in turn, by an annular neutron reflector. The fuel region contained a solution of enriched uranyl nitrate in a mixture of H/sub 2/O and D/sub 2/O and the reflector was a composite of two annuli, the inner one of D/sub 2/O surrounded by one of H/sub 2/O. In most experiments the ends of the assembly were reflected by H/sub 2/O. Important results evaluate the absolute thermal-neutron flux to be expected in the design reactor and describe the flux distributions within this type of assembly. It was also observed that the cadmium ratio along the axis of the assembly was about 100, showing that a highly thermal-neutron flux was truly developed in the trap. It was shown that reduction of the hydrogen …

Report containing experiments conducted at the Oak Ridge National Laboratory's Critical Experiments Facility. These experiments involved "determining the critical parameters of the assembly with varying amounts of neutron absorber in the D20 reflector, measuring the relative neutron flux distribution for various conditions of reflector poison and fissile solution concentration, determining the ratio of the absolute thermal-neutron flux in the center of the assembly to the power in the fissile solution , and determining the effect on reactivity of reducing the hydrogen density in the central region of the assembly" (p. 1).

From abstract: "The preparation of pure molybdenum(IV) bromide was accomplished conveniently by the reaction between molybdenum(III) bromide and liquid bromine at ca. 55°. Molybdenum(IV) bromide was found to be soluble in liquid bromine, and this property was utilized to separate and purify the molybdenum(IV) bromide. The solution of MoBr4 in bromine was a poor electrical conductor. No evidence was found for the formation of MoBr4 from MoBr3 and bromine vapor at 2-4 atm. and temperatures from 180 to 400°. However, MoBr4 was completely decomposed at 110-130°, in vacuo, with formation of MoBr3 and bromine."

Equipment Laboratory and Testing Operation was requested to measure the pressure gradient developed across a nozzle and crossheader assembly used on the rear face during reactor operation. A nozzle and pigtail assembly with pressure taps was installed on process tube 2058, and pressure taps were located at the center, at tube 2058, and at the end of crossheader No. 19-1/2. copper sensing lines were installed between the pressure tape and a tool storage area on zero level, near side of the reactor where our gage could be safely located.

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The objective of the test described in this report is to authorize loading KE Reactor tubes 4396, 4397, 4398, 4496, 4497, 4498, 4596, 4597, and 4598 with 0.947 enriched fuel in support of the X-2 level Battelle Northwest neutron spectrometer, when the spectrometer will be in use.

Using the group vibratibn method of McMurry, the normal frequencies and coordinates of propane and three of its symmetrical deuterium substituted compounds were obtained. The force constants used were taken from a variety of previous works on hydrocarbons. The results give reasonable agreements with the experimental frequency and mode assignments of others. (auth)

This barely legible document contains Zn-65 and Co-60 concentrations and total dose contributions from F area HCR`s. Differentiated include tubes form full rods, tubes form half rods, and outer sheath`s.

A code (RAPTURE) was programmed for the Philco TRANSAC 2000 computer to perform calculations needed for the evaluation of Doppler Coefficients in fast reactors with degraded neuron spectra. This code computes resonance integrals, averages them over resonance parameter distributions, and computes fission and capture cross sections as a function of fuel temperature and of potential scattering cross section per absorber isotope.

A procedure is described for the sequential separation and determination of Arsenic-76 and Phosphorous-32 in reactor effluent water. The analysis can be performed in less than one hour and the arsenic fraction is scanned immediately after separation on a multi-channel gamma spectrometer. The discrete gamma energy of .56 MEV for Arsenic-76 makes further purification unnecessary. The organic phase containing the Phosphorous-32 is dry mounted on a one-inch stainless steel dish for beta counting.

The SNAP-2 Primary System Test loop fabrication was completed with associated flight prototype components including reactor core and boiler mockups for volume and DELTA P simulation, CRU-IIII NaK pump, compact heater, and expansion compensator. A mobile loading system was designed and fabricated with the capability of cleaning the NaK prior to final loop sealing. Loop descriptions, test objectives, and operating procedures are presented. (auth)