The distribution of fuel elements in the EBWR core is described. Spike elements with boron steel strips were provided in the core for burnup allowance. A calibration of the control rod system with Zircaloy followers was made. Void coefficient measurements were made for two different H/sub 3/BO/sub 3/ concentrations. Power runs were made, and reactivity loss vs. power characteristic was obtained. Steam remaining in the core, or carryunder,'' was indicated. Some of the spikes were removed for increasing the power to 80 Mw. Calculations were made of the reactivity loss with coolant void and of the average void vs. reactor power. (D. L.C.)

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

This document details activities of the Irradiation Processing Department during the month of August, 1958. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

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

A letter dated June 3, 1963 to W, E. Johnson, General Manager, Hanford Atomic Products Operation, from J. E. Travis, Manager, Richland Operations Office requested an updated study on the expected life of existing production reactors similar to Sections II, III and IV of document HW-59000. This study covers the ten year period to FY 1973 as the previous study document covered the ten year period FY 1958 to Fy 1968. In Section II, production has been forecast for the existing plant with no improvements beyond the current program for installation of zirconium process tubes in the K Reactors. Section III contains estimates of additional production from the six smaller reactors made possible through overboring the process channels and retubing with zirconium plus the associated water plant expansions. Section IV presents our analysis of the assurance which can reasonably be assigned to the predictions of Sections II and.III. The production forecasts have been prepared from the same base as the current production studies and have been developed within the technical limits listed later in the report,

Attached is the interpolated uranium burnout values Financial Consolidations will apply to the current AEC ten year cost study. We have made additional copies which may be of use to you in future studies since the source data was based on the most recent burnout schedule obtained from Washington AEC. We currently have a program to interpolate future burnout schedules. If you have a need for such a program, please contact the author.

The monthly report for the Hanford Laboratories Operation, May 1963. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, and applied mathematics, and programming operation are discussed.

Reactor water leaks cause outage time for locating and eliminating the leaks and for repairing damage paused by the vater that enters the reactor. The justification for equipment and process changes necessary for reducing the number of leaks and the resultant outage time is dependent on the number of leaks expected. This document presents an estimate of the future reactor water leak frequency and the average outage time charged to each leak. A water leak is defined as any entry of vater into the reactor moderator during operation. Future vater leaks are expected to result primarily from Van Stone flange failures and miscellaneous causes such as mechanical damage, fuel ruptures, and transverse cracks. It is assumed that continuous emphasis will be placed on corrosion monitoring to determine the tubes that must be replaced to prevent leaks because of internal and external corrosion. The number of water leaks is expected to average between 80 and 100 per year in the future. The normal-range of the number of leaks probably will be between 60 and 120 per year as a result of normal variation in the process tube physical condition. The number of outage hours charged per leak is expected to continue …

At the request of Lawrence Radiation Laboratory, four plutonium strips 0.050 inches thick, 0-50 inches in width and 6 inches in length have been examined by metallographic, chemical, X-ray, density and dilatometric methods. The data obtained will help calibrate high energy electrical discharge equipment used to rapidly heat plutonium.

This document describes the detailed test procedures to be used during the initial phases of the physics test program--the approach to critical and minimum critical size tests. During these initial phases, several physics characteristics of the reactor will be evaluated in addition to determination of the critical loading. Overall flux distributions will be determined in the critical pile and used to evaluate the lattice buckling characteristics. Theoretical calculations normalized to experimental values from exponential and PCTR tests have been used for estimating the experimental results for each phase. The startup tests will be used to verify the preliminary calculations and to define the reactor characteristics within the precision required for safe and efficient operation of the reactor. Detailed and accepted test procedures are required for maximum test efficiency and reactor safety during the physics test program and for sufficient overall scheduling into the N-Reactor operational test program. Hazards` analyses have been performed and are presented for cases which are unique to the physics startup tests or significantly differ from the Design Hazards Review.

The research and development activities reported on are: enriched fuel (1.25% U{sup 235}), flex 2 code development, Pu burning and recycle, single tube fuel element, heavy isotope production, U{sup 233} -- Th{sup 232} fuel cycle, physics input for economic studies, Pu{sup 240} effective resonance integral, N-Reactor operator certification lectures, bases for control and nuclear instrumentation, and shield evaluation. (GHH)

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A half-plant low alum treatment test began at F Reactor on January 16, 1963. The test, which had been prompted by results obtained from a statistical analysis of fuel element ledge corrosion attack, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge and groove corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from normal production fuel irradiated in process water treated with two different alum feed rates. This report presents the results from 20 fuel charges, ten from each side of F Reactor, which were discharged such that the near side fuel pieces had been exposed for 75 days to water treated with 18 ppm alum and the far side pieces had been exposed 75 days to water treated with 8 ppm alum.

Recent high tank drawdown data obtained at F reactor during periodic high velocity flushes indicated a reduction in flow to the reactor due to building of corrosion products in the high tank discharge lines. The apparent flow reduction at F reactor was verified by conducting a high tank drawdown test. Visual inspection of B, D, and F high tank discharge lines revealed comparable densities of corrosion product buildup, and high tank drawdown data during high velocity flushes at B and D also indicated a reduction in flow. High tank flow rate to the reactor under simulated emergency conditions needed to be accurately determined to establish current backup adequacy. This report presents the results of the tests which were conducted at D reactor on January 20, 1963 and at B reactor on February 11, 1963.

Over the past few years we have developed a system which allows several different liquid hydrogen and liquid deuterium targets to be placed in the electron beam of the Stanford Mark III accelerator and which can be operated from a remote position. The geometry of our scattering chamber requires that the motion (of the various targets into position) must be in the vertical direction, but that the overall height of the system be constant. These requirements are met by using an internal retractable bellows arrangement to raise and lower the target cells. As many as five cells can be used in conjunction with a single reservoir system.

Criteria provided in this report delineate the objective, bases, and functional requirements that shall govern the preparation of detail design of the bauxite-sulfuric acid feed facilities to be installed in the 183-KE and KW Buildings. These facilities will produce the chemical coagulant used in the treatment of Columbia River water in the water plants and thus replace the existing liquid alum feed systems used for this purpose. The objective of this document is to define the operational and technical requirements of the new process and to outline the functional requirements of the proposed facilities for the purpose of detail design. The criteria below define the requirements for a single K Area water plant. Unless otherwise stated they shall apply for both K Area water plants.

This document is a recommendation to change the MWD factors used by Production Computing Operation, IPD, to calculate Tritium yields from target fuel irradiated in the Hanford Production Reactors. The linear Tritium yield equations in the Nuclear Materials computer programs will not be affected by this change.

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.

Heat transfer calculations were performed with the aid of the IBM 7090 to determine whether or not feasible graphite channel coring patterns could be adopted at the five older Hanford Reactors. The purpose of front and rear process channel coring is to significantly reduce or eliminate net expansion of the fringe graphite by raising the operating temperature above the annealing temperature of 300{degree}C. The results of the study show that such coring patterns are possible. Also, it was found to be possible, and indeed desirable, to standardize the patterns into one front face coring pattern and one rear face coring pattern for all five of the reactors: B, D, F, DR, and H. The resulting coring patterns are presented. These coring patterns will significantly reduce the net rate of expansion in the filler blocks and consequently reduce the inlet and outlet humps in the process channels. This will allow standard 8-inch fuel elements to be charged in all tubes. The afore-mentioned coring patterns will limit the pile gas atmosphere to a range of between 90% He - 10% CO{sub 2} and 100% He. If a greater percentage of CO{sub 2} were used following the adoption of the coring patterns, the …

The conduct of outage work in the discharge areas of the Hanford production reactors has always been complicated by personnel exposure to radiation from the contaminants which accumulate in the effluent water piping in the course of normal operations. These containments represent trace impurities in the cooling water, together with corrosion products, which are irradiated in passing through the reactor and then deposit in the unshielded rear face piping. Radioactive residues also accumulate on the external surfaces of this same complex and congested hardware. The radiation from these depositions increased as power levels and effluent water temperatures were raised, and by 1956 it was apparent that corrective steps would be essential to effective future maintenance work in the discharge areas. It is the purpose of this report to summarize the subsequent development and application of the chemical process for rear face piping decontamination which is being used so effectively in connection with the tube replacement program and other major outrage work.

This report discusses nuclear processes involved in heavy isotope production in the N-Reactor. Production methods, production rates, and the nuclear advantages of the N-Reactor are presented.

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