A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locations in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.

The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10 percent uniform oxidation of graphite results in a loss in strength of approximately 50 per cent. This value of a strength loss of 50 percent has been arbitrarily established as the limit to graphite oxidation over the lifetime of the reactor stack. This limit has been further arbitrarily interpreted as an oxidation rate of not more than two percent per thousand days of reactor operation. At this rate limit, the reactor stack could be operated for 15--20 years before the graphite strength was reduced 50 percent. As with any chemical reaction, the rate of reaction of graphite with oxygen, carbon dioxide, and water vapor increases with increasing temperature. The graphite temperature limits presently in effect at the smaller reactors were therefore primarily established to hold the reaction rate between graphite and carbon dioxide at two percent per thousand operating days and were based on laboratory and in-reactor experiments which are presently believed to have been partially contaminated with oxygen from air leakages. It is therefore the …

The 02 Research and Development program for the Irradiation Processing Department is contained in seven 189-forms, R-1 through R-6 and M-1 and M-2. The subprograms designated by R pertain to reactor research and development, and those designated by M pertain to the metallurgy or the fuels research and development. This report covers the progress in each of the subprograms on an individual basis and follows the format proposed by RLOO. Although the reports are brief, the results are presented in a semitabular form and are essentially complete. Only the activities of the first four months of the fiscal year are included. It should be noted that the over-all IPD program is based on a nearly uniform expenditure of funds over the entire year and the expenditures to date reflect this general pattern. A number of items are listed in the reports on the subprograms that have been performed on product cost funds but were made possible by the research and development effort.

The amount of enriched uranium to support various type of target elements was determined in the Hanford Test Reactor. This report describes the measurement of separate and striped loadings for canned thorium oxide, bismuth, lithium-aluminum, and, in addition, bare depleted uranium. This experiment was undertaken to provide information needed to estimate the E-metal requirements in a production reactor when replacing natural uranium with target material. The amount of enriched uranium which is required to maintain the reactivity in support of target materials must be known to properly predict total reactor reactivity. Danger coefficient measurements were made during original startup tests to provide this information. However, fuel dimensions have changed and new target materials have been introduced which have not been tested in this manner. The target of particular interest at this time of thorium oxide.

This monthly report details activities of the N-Reactor Research and Engineering Department during the month of December 1964.

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This report presents the results of a recent analysis to determine if there is a practical method for using carbon dioxide for pH adjustment at the 100-K Reactors. Carbon dioxide could be recovered from the boiler stack gas and introduced into the K Reactor process water. The approximately $240,000 installation cost would be repaid in about one year by the reduction of sulfuric acid costs. The proposed system would consist of a blower, a purification system, and a distribution system for conducting the gas from the boiler house stacks to spargers in the process water clearwells. The blower would draw gas from the boiler stacks and then blow it successively through the purification system and the distribution system. The purified gas, consisting primarily of carbon dioxide, oxygen, and nitrogen, would bubble up through the water in the clearwells, where the carbon dioxide would be absorbed by direct contact with the water. The development work necessary before the use of carbon dioxide can be recommended consists primarily of in-reactor single tube tests followed by a half-reactor test. These tests are necessary to show the effects of carbon dioxide on aluminum corrosion, effluent activity, and reactor hydraulics. The single tube tests already …

Data in this report are significant to Pile Physics calculations and are submitted by the Area Physicists at the respective reactors. Reactivity in non-uranium charges and in radial and spike enrichment is calculated either by one-group diffusion theory or by the simpler flux-squared weighting method. The former method provides a more accurate evaluation for larger enrichment inventories; the latter method is sufficient for radial enrichment of less than about 0.25 percent excess K.

Each of the 25 monitor tubes at B and D reactors have been gauged four to six times since they were installed in late 1961 or early 1962. Since October, 1962, the WTG data indicates the average corrosion rate, is 0.25 to 0.30 mils per month. The corrosion rate of the thinnest tube is about the same as the average tube. As expected, the corrosion rate is low and believed to be a measure of top-wall corrosion. It is estimated that it will take approximately one more year (at 6.6 pH) before the bottom wells will have the same average minimum wall thickness as the top walls. During this period, the main problem will be to determine whether the top wells will continue corroding at rates between 0.25 to 0.30 mils per month. it is recommended the 25 tubes at B and D reactors be gauged with the WTG every metal cycle during the coming year. The Process Engineer will continue to analyze the WTG data as it accumulates and recommend further monitoring or replacement, subject to the details authorized in the parent test.

The purpose of this test was to demonstrate and characterize the simultaneous generation of plutonium and tritium. The data charged was February 10, 1964 at 12:00 noon. The fuel elements were thirteen 26.1 inch co-product test assemblies. The fuel was exposed for a total of 2,828 hours and at run conditions for 2,288 hours. The date discharged was June 7, 1964. Goal exposure was met.

This report discusses the review of the proposed thorium nitrate specifications,the preparation of tentative product specifications for Purex U-233 and thorium nitrates, and an estimation of the feasibility, cost and product composition for providing U-233 oxide at Hanford. The tentative specifications proposed for U-233 and thorium product for the Purex Plant are presented. The values chosen represent the understanding of feed material and reactor processing needs and the current estimates of Purex Plant capabilities.

Discussed topics on current reactor production fuels include: uranium core production data, uranium specifications, induction heat treating, fuel performance, residual stress measurements, AlSi process development, thoria program, and self-support program. The N-reactor fuels production is discussed. And concerning N-Reactor fuels development, process development and irradiation experience are discussed.

For the test detailed in this report, the charge consisted of three NIEL fuel elements, two lithium-aluminum target elements, and front and rear crud probe train assemblies containing four ceramic fueled crud probes as authorized by PT IP-601. The objectives of these tests were to evaluate the crud forming characteristics of loop coolant containing ammonium hydroxide for pH control by operating at conditions generally regarded to induce high crud concentrations and to provide lithium-aluminum samples which may be used to determine production and extraction information.

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys when loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected …

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The C-1 Loop is a small in-reactor water loop located in the 105-C Building. The loop was originally designed and constructed for nonfueled testing of the effects of reactor radiation on sample corrosion, crud deposition, and coolant radiolysis. The loop was modified in 1964 to permit the irradiation of aluminum-clad fuel elements and subsequently used for about eight months for the in-reactor testing of aluminum-clad, plutonium-aluminum fuel assemblies at high coolant temperatures and pressures. The facility is currently inactive but is available for the performance of special testing. Prior to nuclear operation of the C-1 Loop the mechanical equipment was carefully checked and the backup coolant system was experimentally verified to perform as designed. A safety analysis was also made with results summarized briefly in Reference and in more detail in the production test which authorized nuclear operation of the C-1 Facility. This report updates those earlier documents with improved descriptions of the C-1 Loop and its operation including analyses of postulated accidents.

The first series of tests in the C-1 Loop was designed to test the corrosion characteristics of several aluminum alloys at coolant temperatures up to 290C. Details of the test are documented elsewhere. Four fuel assemblies were irradiated. This report sizes the operating history and the pertinent test data for each of these assemblies.

A research and development project for study of the effects of reactor effluent on Columbia River water quality is being sponsored by the AEC Division of Production. Work was started in October 1962, the first effort being aimed at furnishing an immediate answer to a security question: how closely can Hanford production be estimated by measurement of river temperatures? An interim progress report gave the results of the preliminary investigation. During the calendar year 1963, the study was expanded to meet broader program objectives. This document is a progress report for the year, covering the temperature and effluent distribution phases of the program. Progress on the chemical characteristics phase of the program is documented separately. Figure 1 shows the section of the river under study.

This is the monthly report for the Hanford Laboratories Operation, November 1964. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research.

This report for November 1964, 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; and weapons manufacturing operation.

Objective of the KE reactor test is to determine the hydriding corrosion rate of Zircaloy-2 at high temperature in a damp helium atmosphere, and to determine the outgassing rate of tritium generated in a lithium-aluminum alloy. The irradiation capsule design, operating conditions, etc. are discussed.

This paper contains the minutes of a meeting on Columbia River contamination at Richland, Washington, held on July 16, 1964. The US Public Health Service sponsored the meeting and representatives from the following organizations were present: Washington State Pollution Control, Washington State Department of Health, Richlands Operations Office, AEC Headquarters, and General Electric Company. The meeting dealt with reducing the amount of effluent contributed to the Columbia River, the amount of reduction achievable, the costs involved and the status of technical feasibility studies. This document was declassified on January 26, 1994.

This report details what types and amounts of liquid wastes were disposed in the ground at Hanford in October, 1964.

This document details activities of the irradiation processing department during the month of November, 1964. 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; and Financial Operation.