The following analysis of the induced radioactivity in SNAP-50/SPUR electrical machinery having a high cobalt content is submitted. Induced radioactivity in the flight vehicle will contribute negligibly to allowable radiation levels. This is especially so due to the low neutron to gamma ratio of assumed radiation damage tolerances to semiconductors. A calculation to estimate the order of magnitude of induced radioactivity in cobalt is attached. The calculation is based on a best guess of the neutron spectrum directly behind a lithium hydride shield. The resulting low cobalt activity and associated dose rate of about 1 mr/hr at 10 ft from a generator or a motor is insignificant. Although the evaluation indicates insignificant levels of induced radioactivity, this conclusion is not applicable to a ground test. Neutron moderation and scattering from a containment vessel and biological shield would greatly perturb the neutron environment behind the flight shield. Posttest handling of all components within the vacuum test chamber will undoubtedly be a problem. Notwithstanding the importance of limiting induced radioactivity, other considerations such as economy, cooling and vacuum requirements will largely dictate the final facility design. In summary, an activation analysis involves the overall facility design and will not be readily resolved. …

Report issued by the APDA over studies conducted on Doppler coefficient fast reactors. Calculations, methods, and results of the studies are presented. This report includes tables, and illustrations.

This document details activities of the N-Reactor Department during the month of October 1963.

Interaction of {alpha} particles with elements of low atomic number leads to the emission of neutrons by the ({alpha}, n) reaction. The light element content may become one of the limiting factors in the production of U-233 acceptable for projected uses. Neutron production by the ({alpha}, n) reaction was calculated as functions of U-232 and light element contents.

(VSR = vertical safety rod.) This report summarizes results of development programs which have evaluated vertical rod channel sleeving materials and provided flexible vertical rods, acceptable for both interim use before rod channel sleeving, and for subsequent use in sleeved channels. B{sub 4}C is the rod material; graphite and Al oxide are among the sleeve materials.

The objective of this test is to evaluate the irradiation dimensional stability of uranium fuel cores produced by an induction heat treating process.

Production efficiency studies of the E-N loading require that conversion ratio values for both plutonium and tritium be defined. Changing of the core-loading charge makeup with each core load to obtain a more efficient loading has theoretically resulted in increasing the tritium conversion ratio. Tritium recovery data from the third H Reactor E-N loading has recently become available for analysis. The buildup of product is much slower in fringe blanket material then in core target pieces. The second group of fringe conversion ratio data has only recently been obtained for analysis, the material analyzed was from the DR Reactor, irradiated under conditions closely paralleling those in the H Reactor blanket load. This document reports the conversion efficiencies for tritium production in Hanford E-N core and blanket loadings `indicated by these most recent data.

This document considers the costs of rhenium formation as produced by irradiating tungsten. Two isotopic compositions of tungsten are considered for the study. The cost for reactor-formed rhenium appears to be prohibitively high -- over $20 per gram. This cost would exist for tungsten containing 90% 186, 9, 184, and 1% tungsten 182 and 183. The cost of alloy made fro natural isotopic compositions of tungsten would be higher by a factor of 3, and would take prohibitively long to produce significant quantities of rhenium. Thus, detailed numbers are not shown or considered for the natural isotopic composition of tungsten.

None

Following the startup of 11-6-63 at 105-B, there was an operating period where graphite stringer temperatures were above allowable, and a PCA was authorized for continued operation. It is probable that the combined effects of raising the CO{sub 2} rapidly along with the fast approach to full power level caused the temperature effects. The CO{sub 2} reached 67%. As an interim step, it is suggested that startup CO{sub 2} additions be tailored to level off at 5--10% below the equilibrium level, and that resulting graphite temperatures limit the power level accordingly.

Progress is reported for the month of October, 1963 in the following research and development activities: Lattice parameter and spectral index experiments; T{sup 3} and Np{sup 237} production rates for a highly enriched fuel loading; graphite heat generation; technical bases; physics study of single tube fuel element; thirty-six tube special 1.25 w/o fuel test loading; and code development FLEX 3.

The current E-N demonstration program began during mid-November 1962 in the AlSi Shop. Man meetings and conferences were held prior to this time to determine various methods by which the ultimate quality of the uranium portion of an E-N load could be enhanced at the expense of the material yield, if necessary. As a result of these meetings, it was decided that a superior grade fuel element could be manufactured by raising or lowering, as the cause may be, certain limitations that were currently governing the normal ``F`` process production. The new limitations were presented in Process Work Request 116 (1962), and the program began with fuel lot KY-312-E on November 19, 1962.

This report contains the feasibility and cost data for the production of tungsten-rhenium alloys from tungsten targets in the N-Reactor. The two types of target elements assumed were: (a) tungsten containing 90 a/o tungsten-186, 9 a/o tungsten-184 and 1 a/o tungsten-183 and 182, and (b) tungsten of natural isotopic composition.

Distortion of the graphite stacks of the Hanford IPD reactors due to neutron irradiation effects has grown to where major corrective or compensatory action must be taken to preserve the operability of the safety control systems and life of the reactors. For the past few years, the line has been satisfactorily held by an aggressive maintenance program and short-range solutions. However, longer range, more permanent solutions are needed and high priority is being given to studies toward these ends. Important from a reactor life standpoint (and of more immediate concern and the subject of this discussion) is the jeopardy the reactors face from the use of the Ball-3X system. Permanent loss of balls in cracks in the graphite stacks following a ball drop is a real possibility in most of the reactors today. Recent, detailed investigations of the internal stack conditions in the K Reactors have revealed gaps opening into the vertical ball channels which are as wide as three inches. Any means to measurably reduce the probability of an inadvertent ball drop without compromising reactor safety could be a much cheaper alternative to any presently contemplated solutions which are directed towards physically preventing loss of balls to the graphite …

This document is explanatory in nature and is intended to clarify certain questions about reactor neutron flux spectra in various AEC production facilities. Simplified models are used to illustrate neutron ``temperature,`` spectral ``hardening,`` and the so-called ``Westcott R.``

Feasibility and cost data are supplied for the production of tungsten-rhenium alloys from tungsten targets in the N-Reactor. The two types of target elements assumed were: (a) tungsten containing 90 a/o tungsten-186, 9 a/o tungsten-184 and 1 a/o tungsten-183 and 182, and (b) tungsten of natural isotopic composition (28.4 a/o tungsten-186, 30.6 a/o tungsten-184, 14.4 a/o tungsten 183, and 26.4 tungsten-182). It is assumed that the average thermal neutron capture cross section for the tungsten-186 is 32 barns.

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A half-plant low dichromate-low pH test was started at C Reactor on March 1, 1963. The test will demonstrate whether l.0 ppm sodium dichromate provides adequate corrosion inhibition when,the reactor coolant pH is 6.6. The test monitoring will consist of ex-reactor tube examinations, in-reactor.wall thickness measurements, coupons, visual examination of fuel elements and fuel element weight loss measurements. Previous interim reports have discussed the results obtained from the visual examination of two twenty column discharges obtained as follows: one discharge prior to the start of the test; one discharge such that the fuel was exposed to-coolant treated at both 7.0 and 6.6 pH. This report discusses the-results from.the visual examination and-weight loss measurements on fuel irradiated under test conditions, the results-from coupons-exposed in the downstream spacer pattern, and the ex-reactor examination,of two process tubes.

This report, for October 1963 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.

The development status and incentives for the hot the sizing process for reactor production fuels is reviewed and updated and economic justification for the process for four different case studies presented. Based on whether the end bonding step can be eliminated, payout periods of 1.2 to 1.7 years can be realized if the present AlSi process (two-shift, five-day week) is converted to hot die sizing for all reactors. For K-Reactor operation only, Payouts of 1.5 to 2.2 years can be realized by converting to hot the sizing. Should incentives exist for the present six-inch and eight-inch length fuel to longer models (10, 12 inch), costs would be increased in the AlSi process, which-would not be incurred in hot the sizing. This provides additional incentives for hot the sizing and reduces all payout periods by about 0.2 years.

PITA-IP22, SUPP II has pointed out that distinct processing advantages may be realized by simplifying the E-N core by minimizing the number of different kinds of charges and the number of target slugs per charge. In an effort to experimentally verify the theoretical effects of converting to a simplified loading, a 102-tube test section, consisting of three 5.0 inch Li-Al target elements per tube was recently charged in H Reactor, under authority of the PITA supplement. The purpose of this document is to evaluate the basic reactivity of the 102-tube test section, and to determine, from an operational physics viewpoint, whether full scale operation with the simplified core loading would be feasible.

This document reports the result of an engineering analysis and evaluation of the Zone Temperature Monitoring System. The main function of the system is to protect the reactor from localized increases in reactor power which would be reflected by increases in the tube outlet temperatures in the affected zones of the reactor. This function has been implemented by the installation of immersion temperature detectors on 109 representative process tubes. The signals from the detectors will be monitored and will automatically annunciate, initiate a power setback, or initiate a reactor scram should the outlet temperature of one, two, or three tubes exceed an adjustable, preset limit.

Both the inner and outer tubes of the N-Reactor fuel elements will have self supports spot welded to the lateral heat-transfer surface of the element. A crevice a few mils thick will exist around the weld between the support tab and the cladding. Because of the heat flux through the cladding at this point and the insulating effect of the support tab, the temperature in this crevice will be higher than that on the free surface away from the support. This can result in boiling in the crevice leading to concentration of LiOH (or impurities in the water) to a level where it can cause severe corrosion of the Zircaloy-2 cladding. The tests described in this report were conducted to determine whether such attack might be encountered in N-Reactor.

The production of Po{sup 210} by irradiating Bi{sup 209} has been proposed as a possible use of N Reactor. The purpose of this study was to make an estimate of the production rates for various exposure times and flux levels. The production rate was calculated for the three flux levels of: (1) 5.0 {times} 10{sup 13} neutrons/cm{sup 2}/second, which is about the average flux level expected in the N Reactor graphite moderator; (2) 6.74 {times} 10{sup 13} neutrons/cm{sup 2}/second, which is the estimated flux level in the 68 empty process tubes on the top and bottom of the N Reactor core; (3) 7.5 {times} 10{sup 13} neutrons/cm{sup 2}/second, which is the estimated flux level present in a flux trap made by removing a column of fuel. In addition to the above calculations, an estimate was made of the fuel enrichment needed to support a column of bismuth in the core of the reactor; and the percentage loss in the Pu production was calculated for the reactor.