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.

The objective of this test authorization is to increase reactivity and thus reduce short term enrichment requirements by increasing graphite temperature limits during low exposure operation following full central zone discharge.

This document presents experimental data obtained during a series of tests which were completed at 105-D and DR Reactors in February and March, 1960. No analysis of the data is included in this document. The tests were: PT-301-I, II -- Reactor cold, full flow, BPA power failure; PT-278-A, III B -- 1170 MW, full flow, BPA power failure; and PT-278-A, II -- 1190 MW, full flow, poison push causing bulk surge and scram.

A study was made to determine the optimum arrangement of the fringe flow zones at KE Reactor. Also considered, was the possibility of converting part or all of the existing fringe zone (actually three flow zones) from solid metal to I & E metal to decrease rupture potential in these low flow zones. The necessity for this study was indicated by high tube outlet temperatures in the fringe flow zones and the recent occurrence of two solid metal ruptures in process tubes located in the fringe zone of the reactor. Two additional solid ruptures occurred during the period that conversion to I & E in zone 2, as recommended below, was being completed.

As power levels have been raised at the K Reactors, the graphite temperature limit has caused significant production losses. These losses have come from three sources: (1) Occasions vhen pile power was directly limited because the graphite temperatures were limiting; (2) Losses in ECT because control rod movements to control graphite temperatures are not necessarily the best rod movements for pile, flattening; and (3) Losses due to lack of control in the final stages of an operating period because COp could not be used as a means of controlling long-term reactivity gains. This document attempts to establish the magnitude of these losses and shows the justification for increasing the graphite temperature limit. The data which is presented here is based upon operating data from December, 1959) through March, 1960, at KW. The results apply to KE as well as KW.

The purpose of this test is threehold; to test the workability of the flapper and/or ball cap -- actuator combination under reactor operating conditions to determine the effect on the K Reactor operating parameters of the rapid discharge of a column of slugs; and to ascertain the radiation levels present at various locations in the 105 Building when a column of irradiated slugs is discharged during operation.

The intent of this document is to establish criteria for equipment for: Receiving, handling and component inspection; miscellaneous handling and storage; final inspection; and finished storage in the new fuel cladding facility.

None

None

A compilation is presented of results obtained to date from a number of reactor physics calculations for the molten salt reactor experiment (MSRE). Included are one-dimensional multigroup and two-dimensional twogroup calculations of critical mass, flux, and power density distributions; gamma heating in the core can, reactor vessel, and core support grid; drain tank criticality; and an estimate of the beta , gamma , and delayed neutron dose rates due to fission products in the fuel contained in the pump bowl. For a cylindrical core 54 in. in diameter and 66 in. high, graphite-mcderated with 8 vol% fuel salt, the calculated critical loading is 0.76 mole% uranium (93.3% U/sup 235/), which is equivalent to a critical mass of 16 kg. At a reactor power of 10 mw, the peak power density in the core assuming a homogeneous mixure of fuel salt and graphite is 10 watts/cm/sup 3/, the average power density is 4 watts/cm/sup 3/. The computed peak thermal flux is 7.3 x 10/sup 13/ neutrons/cm/sup 2/ sec and the average is 2.5 x l0/sup 13/ neutrons/cm/sup 2/ sec. Gamma heating prcduces a power density of 0.2 watts/cm/sup 3/ in the core wall at the midplane and 0.4 watts/cm/sup 3/ in …

In developing a method of numerical analysis for the solution of thermal- stress problems special emphasis was given to fuel elements with internal coolant channels. Numerical techniques ior reducing the partial differential equation system te a form suitable for numerical solution and a new iteratlve method of solving large systems of linear algebraic equations were employed. Computer codes were devised to obtain the numerical solution of the thermal-stress problems and were used to obtain numerical results for single-hole and seven-hole hexagonal elements and plate-type elements. Comparisons were made between analytical results and numerical results for the case of t:.ie simple annulus shape. (auth)

None

Low attack rates (0.1 to 0.5 mpy) were displayed by coupon specimens of type 347 stainless steel, titanium RC-55, and Zircaloy-2 which were exposed for 2877.5 hr in an oxygenated slurry of Th--8% U oxide, 116.5 hr in water, 6.9 hr in 5% HNO/sub 3/, and 4.3 hr in 3% triscdium phosphate during mns SM-6 through SM-9 in the slurry blanket mockup. The leading coupon of type 347 stainless steel showed a slightly higher rate than the other stainless steel ccupons due to entrance effects. Specimens of SA-212-B carbon steel displayed average attack rates of 2.9 mpy. (auth)

Wood, coke, and coconut-shell charcoals were evaluated for fluorine entrapment. The coconut-shell charcoal produced the smallest amount of solid and liquid reaction products. Efficient removal of fluorine was accomplished by the coconut-shell charcoal in a 5-in.-diameter reactor with a feed containing 25% fluorine at flow rates from 100 to 400 scfh and reactor-wall temperatures of 1200 to 1800 deg F. (C.J.G.)

The SNAP II system consists of a reactor heat source, a mercury Rankine engine, and an alternator. The problems involved in selecting materials for the SNAP II mercury system were studied. A discussion is given of the corrosion mechanisms involved in a system in which mercury is the working fluid. The problem resolves itself into selecting materials with the best combination of engineering properties for the application and highest resistance to mercury corrosion at the anticipated temperature. (auth)