This report, from the Chemical Processing Department at HAPO for November 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

The results of a multi-group, diffusion type code computation for the fast neutron flux distribution in C, K, and N reactors are presented. Relative Ni activations in conjunction with the calculated neutron spectra are used to obtain values for the integrated fast flux per MWD/AT for C and K reactors. The relative, theoretical, integrated fast flux values at the same adjacent fuel powers in C, K, and N reactors in the mid-plane of the filler layers are: (1) C = 1.0; (2) K = 1.3; and (3) N = 2.6. For C reactor, the best estimate of the integrated fast flux per MWD/AT is: 5.0 {times} 10{sup 16} nvt (> 1 MeV) = 1 MWD/AT.

The iron thermal shields of the Hanford reactors are cooled by means of water flow through thermal cooling tubes embedded in the shield blocks. The flow rate, temperature rise, and allowable pressure in the tubes and the conditions under which some of the tubes may be out of service are specified in the process standards. Shield heat formation and heat transfer calculations are necessarily based on broad assumptions and are therefore usually reliable only for order of magnitude and trend predictions. In order to specify condition of shield cooling under which the reactors may be safely and economically operated data must be available regarding operating temperature as related to the flow of cooling water through the shield. Some of the data on which the current standards are based have been determined by extrapolation to present conditions of measurements taken several years ago. The purpose of this test was to establish current data that may be used in updating the thermal shield coolant standards. Measurements were taken of the operating temperatures experienced in an uncooled thermal shield cooling tube in relation to the specific power of the adjacent process tube. Conditions were varied by adjusting the flow in the thermal shield …

The object of this investigation was to determine an economical method of preparing pure lithium hydroxide from a mercury-contaminated lithium hydroxide monohydrate salt presently produced as a waste product. Pure lithium hydroxide has application for chemical removal of carbon dioxide from air and general open market sale if the mercury contamination is reduced to approximately one part per billion. Because of the uncertainty of the form of the mercury contaminant, different purification methods were explored on a laboratory scale which could be applied to the industrial waste stream. The experimental results indicate that the predominant mercury contaminant existed as mercuric oxide, which was deposited in the by-product salt when the solubility of mercuric oxide, 60 ppm, was exceeded in aqueous lithium hydroxide solution. To purify a fraction of the industrial by-product salt, a crystallization system, utilizing the difference in solubility of lithium hydroxide and mercuric oxide, is proposed. Total stream purification, using sulfide treatment, is expected to be less effective than recrystallization due to the difficulty in physical removal of the mercury contaminant, as mercuric sulfide, from solution.

To permit the irradiation of one dummy fuel element assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

This document details activities of the Hanford Laboratories Operation for the month of August 1960.

Experimental data on plutonium yield and U{sup 235} burnout are now available on the E metal portion of three central zone striped E-N columns and one fringe blanket E column. These data and an overall E-N load conversion ratio based on experimental data are now reported.

This report presents a proposed test program to study the corrosion of aluminum-clad fuel elements under conditions expected at the more severe conditions which will be encountered in operating the present Hanford reactor at higher power levels.

In 1955 the Fuels Development Operation began irradiation testing of fuel elements in high temperature water. It was assumed that if a new reactor were built at Hanford, it would be cooled by high-temperature, pressurized water. Corrosion tests showed that aluminum-clad production fuel elements could not be used in high-temperature water. Therefore, while work to improve the resistance of aluminum to high-temperature water proceeded, the Fuel Design Operation began irradiation of stainless steel- and Zircaloy-2-clad fuel elements. During 1956 and 1957, stainless steel-clad elements were tested in the Materials Testing Reactor (MTR), Hanford H Reactor Loop, and the KE Reactor Recirculating (KER) Loops. During 1957, a coextrusion method for cladding uranium rods with Zircaloy-2 was developed. The first irradiation of Zircaloy-2-clad fuel from an off-site supplier began in late 1958. The objective of the irradiation was to study the dimensional stability of the fuel rods and a seven-rod fuel assembly. Two coextruded, seven-rod elements were irradiated in KER Loop l.

This report discuss the design of the Plutonium Reclamation Facility the capacity of which will be 300 kilograms per month or 3600 kilograms per year or plutonium. The subject facility, as the name implies, must be extremely flexible in its ability to handle a wide variety of feed materials. The new facility will be operating on a three-shift day, five-day week, 40-week year with an overall efficiency of 75 percent; twelve weeks per year will be required for ``turnaround`` time to enable campaign operation for segregation of feed plutonium by isotopic content.

The proposed increased water capacity for 100-K plants will increase the electrical load to be supplied. The load study showed that the capacity of the existing 13.8 kV system is adequate to carry the increased loads proposed for Project CG-775, while for the 5 kV system, an expanded power system is proposed. Likewise, the voltage regulation on the kV system bus will be excessive, and voltage regulators should be added.

Failure of reactor rear face connectors is a problem at the reactors. The vibration has been studied at H and F Reactors to assist in the development of a permanent replacement for these connectors. Vibration data were otained from rear nozzles, connectors, and crossheaders within the process tube pattern for the following operating conditions: full cold water flow prior to reactor startup, normal operating power level, and during transition from shutdown to normal operating power level. The vibration patterns at 105-F and 105-H were similar in magnitude. Oscillograph data are presented.

This document discusses the test program of evaluating the temperature effect of crud film build-up on fuel element heat generating surfaces in a carbon steel system. This program has three phases: Measurement of the effect of film build-up during normal high pH equilibrium operation; measurement of the temperature effect of film build-up subsequent to a loop decontamination; and measurement of the effect of film build-up in the event of loss of pH control.

This report on fuel element handling presents in some detail the current status of an engineering study which has been underway for some time, and which is continuing. The study was undertaken to determine if it is feasible, and if it is practicable, to revise the method and equipment used for fuel element handling with existing charging machines.

The purpose of this report is to define the extent of modifications necessary to increase capacities of the 100-B, C, D, DR, F, and H water plants for reactor flows of 90,000 95,000 105,000 and 115,000 GPM, and to provide supporting data for budget studies for increased production.

To date a great deal of emphasis has been placed on flattening the side-to-side and top-to-bottom flux distribution with only minor effort to improve the front-to-rear distribution. Minor variations in the front-to-rear distribution have been achieved by horizontal control rod and Supplemental control positioning. It has-been reasonably well established that the rupture potential for one tube charge increases markedly with higher specific power and temperature; thus there is a great deal of incentive to flatten in the front-to-rear dimension. Although flattening in this dimension will caure increased neutron leakage out of the reactor, this is compensated by increased conversion efficiency resulting from a more uniform exposure distribution within the tube charge. The purpose of this document is to describe the basic analytical methods and the techniques, of flattening front-to-rear through the integrated use of enrichment and poison material in combination with natural uranium, and to point out the requirements to insure that total control criteria is satisfied in the event of a water loss with this loading. For the purpose of this survey report an old reactor, 32-piece charge length, and a symmetrical front-to-rear distribution were considered; however, the methods given can be extended quite easily to different length and …

In response to specific requests of the AEC and as part of a Combined Operations over-all review, a number of engineering studies have been made of alternative methods for increasing availabiltiy of neptunium at Hanford. The report updates an earlier study in which recycling of both natural and enriched uranium was considered for Hanford. The earlier study showed that recycled natural uranium would provide appreciable gains in neptunium availability but at an excessive cost. Recycle of the slightly enriched uranium streams proved a more economical means of realizing smaller but still significant gains in neptunium production. Subsequent to the earlier report, a feasible and immediately applicable scheme for UFI blending has been conceived demonstrating further advantages for recycling the enriched uranium. Approximately 34 kilograins of supplementary neptunium could be produced at Hanford during the next seven to eight years by upgrading irradiated E-metal and NPR uranium through a blending operation at Oak Ridge (rather than in the diffusion cascades) and then recycling the material through the Hanford reactors. (cf Table 1) Such a scheme would conserve uranium-236 for use as a source of neptunium-237 in the reactors without incurring major capital costs. Oak Ridge Operations personnel have estimated that capital …

It is recommended that the ``square pile`` concept be adopted for all disaster total control calculations, and that the basic reactor constants listed in HW-62884, except for Ball 3X local strength at the DR Reactor, be used in applying this method. Curves are included for each reactor type, indicating allowable enrichment based on appropriate local control strengths. (The reactors whose operating methods are affected by disaster total control requirements are B, D, F, and DR Reactors; the remaining piles have sufficient geometrical coverage). An example of the analytical method is included.

Recent borescoping of the KER-1 tube revealed several scratches, pits, and gall marks on the internal wall of the tube. These deformations could limit the operating temperature and pressure of KER-1. This report is a summary of operating history and is compiles to assist in determining what contributed to the condition of the tube.

This report is a compilation of available data on ruptures that have occurred since KER start-up. Some of the data presented in this report are inconclusive, but are reported to insure a complete summary of all potential and confirmed ruptures.

The objective of this test is to determine the DR-Reactor effluent systems characteristics under 95 degrees Celsius bulk temperature operation. This proposed bulk temperature increase from 93.5 to 95 degrees represents a 33% decrease in the bulk temperature suppression below the boiling point. A major aim of this test will be to evaluate the degree of increased maintenance at this higher temperature operation. The basis and justification, test preparation and instrumentation, procedure, costs, outage time, hazards, standards, and responsibilities are discussed in this document.

Use of X-8001 Al alloy as cladding for Hanford reactors was initiated because of superior (laboratory) resistance to intergranular corrosion over that of C-64 alloy. However, since severe pitting attack was observed intermittently, an evaluation was carried out on X-8001 alloy fuel element cladding.

The loop was charged October 30, 1959 with seven 12-inch Natural Uranium Zircaloy-2 clad 7-rod clusters. The test was primarily for the new hot-headed method of end closure used on these elements. The loop was pressure tested at 4000 psi after it was charged. Initially the loop was held at low-temperature to study the buildup of oxygen (presumably from radiolytic decomposition of water). After startup the neutron activity held at slightly above normal but the strainer gamma activity was exceptionally low. Frequent additions of LiOH bombs were necessary to maintain the pH at 10.0 (previously it was 4--5 pH but was raised to pH 10 for this test.) After the temperature was raised to operating conditions the pH held nicely at 10. On November 16, 1959, the heat exchanger exit temperature thermocouple blew out resulting in depressurization of the loop and a reactor scram. Repairs were made during the outage and the loop was returned to normal operation. Eleven scrams occurred during the test period caused by 105-KE or 1706-KER.

The objective of this production test is to evaluate the behavior during irradiation of nominally 1.739 inch OD., 1.074 inch ID., Zircaloy-2 jacketed single tube fuel elements with brazed end closures at operating conditions somewhat more severe than those expected for N Reactor fuel element outer tubes. Enriched single tube elements with brazed end closures and modified fuel element supports are authorized for irradiation in the KEG loops to an exposure no higher than 4000 MWD/T. A typical loading and set of operating conditions are given; these may be modified, however, by obtaining appropriate approvals.