Nucleon-meson cascade calculations were carried out and the dose as a function of depth was obtained for an 800-Mev proton beam incident on a shield. The physical properties used for the shielding medium are only a rough approximation to the properties of any particular medium. Muon, neutron, pion, and proton dose rates and fluxes are listed. (auth)

In a previous report nucleon-meson cascade calculations were carried out for several cases of interest in the design of the transverse shield for the proposed 45-Bev linear electron accelerator at Starford University. In this report results are given for several additional cases. Muon, neutron, pion, and proton intensities as function of energy and distance for varying angles are given. Corresponding doses are also included. (D.C.W.)

S>Critical studies of two fast reactor cores are described: one contains uranium and steel; the other contains uranium, steel, and sodium. Experimental results are given for fission ratio, central and edge reactivity coefficients, fuel bunching, average prompt neutron lifetime, and distributed worth measurements. (auth)

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The Large Component Test Loop is a facility for ASCR providing experimental capacity for subjecting large components, such as moderator assemblies and control rods, to thermal gradients and transients at varying rates of sodium flow to simulate reactor operating conditions. Two separate loops are used for heating and cooling the sodium as the testing is performed. The 3-inch pump loop, modified to operate at 1200 deg F, was operated at a flow of 130 gpm and a head of 25 feet. The 6-inch pump loop, limited to 1000 deg F operation by materials of construction, was operated at a flow of 2000 gpm and head of 65 feet. The rest of the facility has operated satisfactorily at 1200 deg F. (auth)

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The differential cross sections for elastic scattering of 14-Mev neutrons from nitrogen and oxygen were measured in the angular range from 17 to 140 deg , using liquid targets and annular ring geometry. Inelastic scattering to the 2.31 and 3.95-Mev levels in N/sup 14/ and to the levels near 6 and 7 Mev in O/sup 16/ was investigated over a more restricted angular range. The 2.31-Mev level in N/sup 14/ was not excited appreciably, in disagreement with previously reported results. For both elements, the elastic scattering cross sections (determined to an accuracy of 10%) were found to be appreciably higher than the optical model predictions by Rjorklund and Fernbach for scatterirg angles larger than 70 deg . The inelastic cross sections measured are, within experimental accuracy, the same as the corresponding (p,p') cross sections. (auth)

About 298,000 OIIIN aluminum components were received from Harvey for evaluation after major steps were take in the vendor`s plant to upgrade product quality. This material met HAPO specifications and was comparable to Alcoa control components. Finished fuel elements (non-bumper and bumper) canned in the test and control components are tabulated by uranium core type and canning month. Non-destructive and destructive test data did not show any differences in clad fuel elements attributable to component quality. Irradiation status of non-bumper and bumper fuel elements clad in Harvey improved components as of Dec. 31, 1962, is tabulated, as is a summary of ruptures. It was recommended that Harvey Aluminum Co. be allowed to bit competitively on all future Al component orders except for the KVN and KVE component models.

Information concerning the corrosive effects of water in power reactor moderator-coolant systems is presented. The information is based on investigations reported in the unclassified literature believed to be fairly complete to 1959, but less complete since then. The material is presented in sections on water decomposition, water chemistry, materials corrosion, corrosion product deposits, and radioactivity. It is noted that the report is presented as a part of a continuing program in development of less expensive materials for use in reactors. (J.R.D.)

The Doppler coefficient in thorium dispersed in graphite was measured in a reactor spectrum developed in the HTGR critical assembly, using a reactivity technique that compares the reactivity of a single cold element and a single hot element heated to 700 deg K. The considerable correction due to the small perturbation in neutron temperature resulting from the heating has been experimentally determined by auxiliary measurements. An activation technique has been used to check the results. In this activation technique, the thermal-base effect has also been experimentally subtracted by measuring the 1/v component of the captures at each temperature in vanadium. Measurements of the Doppler coefficient obtained by these two different techniques agree to within 10%. This agreement, demonstrated by applying both techniques for the first time in the same laboratory, shows that the previously reported spread of results and apparent disagreement between results of these two techniques are avoidable. The average Doppler coefficient for the temperature interval from 300 to 700 deg K, for a carbon-to-thorium ratio of about 50 in the thorium-bearing region, was measured to be 3.6 x 10/sup -4// deg C to within about 10% uncertainty. The measured results are in satisfactory agreement with calculated values. (auth)

The objectives of this production test are to confirm the KVNS fuel element design, determine the effect of support height on annulus coolant temperature distribution (R Value), evaluate the effect of eccentricity on KVNS fuel element irradiation behavior.

A half-plant low alum water treatment test began at F Reactor on January 16, 1963 at startup from the scheduled January 3 tube replacement outage. The test, which was prompted by results obtained from a statistical analysis of fuel ledge corrosion attack, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from normal production fuel irradiated in two different alum treated process waters. This report discusses the results obtained from twenty fuel charges, ten from each side of F Reactor, which were discharged prior to the reduction in alum feed to establish the pre-test corrosion environment.

This study was undertaken to answer questions asked regarding the required rod stroke for control of modified Tory II-C reactors. All problems described were solved with the Angie code and based on Tory-II-C design problems, RZ 501 and RZ 502, representing hot and cold reactors respectively.

The diffusion of lithium in aluminum was measured at various temperatures with diffusion couples of aluminum-LiAl. The activation energy, E, is 33.3 kcal/mol, and the diffusion factor, Do, is 4.5 cm{sup2}/sec. (auth)

This document summarizes the graphite treatment studies conducted under Design Test 1136 in connection with the retubing of the K Reactors with zirconium tubes.

Experimental work consisted of one new run started and completed this quarter, and the completion of one of the three continuing runs. Of the scheduled operating time, the beam was on for 69.4% of the time, 2.3% of the time was used for experimental setup, and equipment outage took 29.3% of the time. There were two scheduled and two impromptu shutdowns. During one of the scheduled shutdowns the external-beam extraction magnets were installed in the east and south tangent tanks. The other scheduled shutdown was to readjust the Bevatron magnet elevation to correct for foundation subsidence. Internal magnets were also installed. In the new linac development program the ion source was run at 480 kv with a beam current of 100 ma. The linac tank was partially deplated to provide a clean copper surface, and welds and holes were plated with copper. The r-f losses were thereby reduced 20%. (auth)

This document has been prepared for internal use by the General Electric Company to serve as a program for evaluating reactor instrumentation and safety circuit equipment needs. It is intended that this document be used as a guide for defining, planning and scheduling engineering effort; budgeting of capital money; and project planning for new instrumentation systems. Effort will be made to periodically evaluate the status of the programs presented and provide updating information accordingly.After a plant has been built and operated for a number of years, it becomes apparent to operating and engineering personnel that certain modifications in controls and monitoring systems would provide both tangible and intangible benefits. Systems which were once thought to be the primary points of control shift in importance as others become recognized. As time passes this shifting spreads the main control focus from the central control desk to various other areas in the control room. Production rate increases cause instrument ranges and scales to be changed so that information on the process can still be obtained from existing equipment. Response times, sensitivity, limits, and time constants which were figured for one level must be used or revised for new levels. Further, it is discovered …

Critical studies of a simplified mockup of the second loading (Core B) for the Enrico Fermi fast reactor were made. The core consists of a fully enriched, UO/sub 2/-stainless steel cermet clad in stainless steel and cooled by sodium. The high critical mass found for the reactor mockup led to a series of experiments on reducing neutron leakage from the core by surrounding it with different reflectors. With a nickel reflector in place, measurements were made of the reactivity worths of various engineering features of the Fermi reactor, distributed and local reactivity coefficients, various fission and capture rates and fission ratios, and the prompt-neutron lifetime. Finally, with a nickel oxide reflector in place, some of the above measurements were repeated. (auth)

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Unclassified technical reports and papers in the hightemperature liquid- metal field are reviewed. All phases of liquid-metal technology pertinent to the space power development program are covered, including materials development, corrosion, heat transfer, fluid dynamics, instrumentation, component development, physical properties, and power systems. Twelve reviews are included. (M.C.G.)

TRANSURANIUM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING AUGUST 31, 1962 On Process Development, Equipment Development, HFIR Target Fabrication Process Development, Design Studies, Materials Evaluation Studies, TRU Corrosion Studies, Analytical Development and Uranium-232 Preparation. (auth)

MIST is an IBM 7090 program prepared at Brookhaven National Laboratory as part of a cooperative effort for bubble chamber data processing. The program reads the prepared data and outputs it on magnetic tape in a suitable format for input to the Haze-Fog-Cloudy-Fair Programs used with a flying spot digitizer. MIST deletes erroneous data, edits, sequences, and supplies additional physical data not digitized. MIST also prepares detailed summaries and statistics for the events processed and checks digitizer operation. (H.G.G.)

A half-plant high flocculation pH test began at B Reactor on September 24, 1963. The purpose of the test was to determine whether operation of the water plant flocculation basins and filters at a pH higher than 7.0 with subsequent coolant pH adjustment to 6.6 is beneficial as far as radioactivity in the effluent is concerned. It is planned to increase the flocculation basin pH in steps to a maximum of 7.6. Although primarily an effluent activity test, fuel discharged after a metal cycle at each water plant condition will be monitored by visual examination. Thus, it will be possible to determine whether the incidence of localized corrosion is influenced by the various modes of operation. One 20 column fuel discharge, ten columns from each side of the reactor, was obtained prior to the start of the test. This report discusses the results obtained from the first discharge of fuel under test conditions.

This report details test plans for production test IP-560-A. The objective of which is to determine the corrosion effects resulting from reducing the amount of sodium dichromate corrosion inhibitor added to process water from 1.8 to 1.0 ppm at a process water pH of 6.6.