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 …

This report discusses the use of liquid ammonia as a tracer for the detection of cooling water leaks into the piles. It is both safe and cheap and can be detected by methods adaptable to direct-reading instrumentation on a continuous-flow sample. Moderate capital costs and materials costs of less than $50 per pile test are anticipated.

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The objective of this test is to determine, through in-reactor testing, the resistance to corrosion of nickel plated fuel elements, plated by a chemical deposition technique. This program may eventually be composed of the following portions, the first of which is authorized by this test: (1) Irradiate ten columns of 0.5 mil chemically nickel-plated C-64 OIIN fuel elements having various heat treatments, each alternated with x-8001 control pieces. Two columns will be exposed to 400 MWD/T and eight to 800 NWD/T. (2) Should results from the above test be encouraging, separate authorization for the following tests may be requested: (a) Irradiate four columns of nickel-plated fuel and four of X-8001 clad fuel, both groups having purposely cocked pieces, until two ruptures are sustained in each group or until a factor of improvements of 400 is demonstrated at the 95% confidence level, unless the nickel-plated elements fail first. (b) Irradiate five columns of chemically nickel-plated C-64 clad OIIN fuel elements, alternating 0.2 mil and 0.5 mil plate, heat treated at 300{degrees}C for six hours. Two columns will be discharged at 400 MWD/T and three columns at 800 MWD/T exposure. (c) Irradiate approximately thirty columns of nickel-plated, C-64 alloy clad fuel elements …

Because of difficulties in charging eight-inch fuel elements in bowed process tubes in the upper parts of the old reactors, IPD has requested that four-inch I & E fuel elements be provided for these tubes early in CY1961. Approximately 30,000 fuel elements of this design, Model OVN, will be required per year. Since most of these fuel elements will be charged in the upper fringe zones, they will be in-reactor for an extended period of time. For this reason, the quality of this material should be as high as possible. This report contains a summary of tests made to determine optimum canning conditions and to establish process specifications for canning four-inch I & E fuel elements.

This production test is designed to demonstrate that the K Reactors can be operated. with a higher graphite temperature limit than stipulated in present standards without a significant increase in the rate of either burnout or contraction of the graphite moderator stack. It is intended that the increase in allowable maximum graphite temperature will be utilized to permit operation with slightly higher reactor power levels with substantially the same enrichment level of the total uranium fuel charge as is currently in practice at the K Reactors.

This report addresses the critical mass studies, part X.

The outlet water temperature limits for the Hanford production reactors are based, among other things, on the ability to prevent excessive fuel temperatures during or following inadvertent flow losses or power surges. The analyses involved in the determination of the limits are dependent upon laboratory heat transfer experiments which closely simulate actual reactor conditions in a single process tube. The heat transfer experiments are quite complex in that they involve the recording of rapid changes in pressure, temperature, flow, and heat generation rates. As a preliminary to these transient heat transfer experiments, selected steady state experiments are performed to better plan the experimental program and to aid in the analysis of the problem. This document presents the results of some steady state experiments performed at low tube powers for K-reactor. The results are reported separately from the transient experiments because the data are useful for reactor problems involving shutdown cooling and are also of value to people interested in pressure drop of two phase mixtures.

The purpose of this test is to establish the reactor hydraulic flow vs pressure demand curve in the low-regions of around 4--10,000 gpm.

The purpose of this report is to describe orifice combinations which can be utilized in Zone 2, 3, and 4 of the K Reactors while using K-IV I&E fuel elements and to present working data to describe the calibration and.critical flow characteristics of initial prototypes of these orifice combinations. A discussion is included concerning the minimum flow rates attainable by front face orificing while avoiding critical flow conditions.

A study was made to establish the validity of using homogeneous-lattice machine program calculations of the relative conversion ratio of the overbored lattice as compared to the present case. It is possible to draw both qualitative and quantitative conclusions from hand calculations utilizing simplified geometries; it is the purpose of this document to summarize such a study. Numbers included herein are not intended to represent absolute quantities, since the required calculations lie outside the scope of this study. Nevertheless, it is felt that a comparison of the overbored to the existing lattice based on the values presented here is a good ``first order`` approximation.

A brief study of the nitric acid leaching of sheared Zircaloy and stainless-steel-clad UO/sub 2/ fuels was carried out. The study was intended to demonstrate the leaching process and the use of a removable dissolver canister for transport and charge duty. In addition, the effect of sheared element length on leaching time cycles was studied. (L.T.W.)

Alternator development for the SNAP 1 power conversion system is described. A radial air-gap permanent-magnet 6-pole single-phase 2600-cps alternator, rated at 536 watts, 0.8 power factor, and 115 volts was selected. Discussion of requirements and specifications, design and performance, test facilities, and conclusions are included. (J.R.D.)

A program is reported to investigate a method of stopping sodium leaks at 1200 deg F by reacting the leaking material with a gas such as carbon dioxide or oxygen. Such a plugging mechanism would be useful in preventing shutdowns due to sodium leaks such as those in the shell of a calandria core. Results indicate little likelihood of success for this mechanism above 600 deg F. (J.R.D.)

A survey was made of sealing methods and their adaptability to the requirements of the plenum reflector seal on the HTGR reactor vessel. A leak rate through the seal of 43.8 pounds per hour of helium was established as an acceptable leak. Lap joints of steel to graphite or graphite to graphite with a 100 RMS surface finish met the required leak rate and two basic concepts of this type of seal are illustrated. (auth)

This document details activities of the irradiation processing department during the month of May, 1960. 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.

Production of Pu nitrate from separations plants during May was below forecast. A Np recovery campaign in Purex yielded 1.5 kg. Production and shipments of UO{sub 3} met schedules. Unfabricated Pu metal production was below forecast, but all shipments were on schedule. Decontamination efficiency was low in Purex solvent extraction around the time of the Np recovery. The damaged Redox B-2 dissolver is being restored; processing of enriched metal in A and C dissolvers was continued. A spectrograph for inclusions in Pu metal was installed. 4 kg Pu oxide was produced in a continuous direct calciner. Scope design on Purex Np recovery and purification facilities was completed. Other design and contracts are discussed.

The daily check list, the startup procedure both for in-pool and in-air operation, and the normal shutdown procedure for the Tower Shielding Reactor-II are outlined. The sequence of events which leads to each annunciator warning and the automatic corrective action which is then initiated by the control circuits are described. (auth)

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Investigations of materials for use in connection with the SNAP I mercury Rankine cycle power conversion system are discussed. Test programs are outlined and results are tabulated for each candidate material. Several nonmetallic materials and processing procedures were developed which enabled uncooled high-performance electric machinery to operate at 550 deg F in mercury vapor. (J.R.D.)

With the exception of a brief period of slightly elevated chloride level in the secondary blowdown, water-chemistry conditions during the period were satisfactory. During the period, the reactor was shut down for end-of-core-life testing and rearrangement. A set of specifications covering all electronic and electromechanical mechanisms required to control the SM-1 reactor through the rod- drive motors and clutches was prepared and issued. Installation of instrumentation for plant response and system performance was virtually completed. Work on the interpretation of long-lived radiochemical data obtained at the SM-1 during core lifetime was continued. Analysis of all fissionproduct data collected during Core I life has started. Thirty-eight stationary and seven control subassemblies from SM-1 Core II were checked for alpha contamination by a gas-flow proportional-counting technique. The work on the final design of a waste-disposal system for SM-lA was stopped and an investigation of an interim system containing a bypass sampling system was undertaken. Work continued on tests 202, 203, and 204 in the activitybuildup phase of Test Series 200. Core- physics measurements were taken at end of Core I life to complete the series of measurements made throughout the lifetime of the core. (W.L.H.)

One hundred and seventy five grams of americium in a hydrochloric acid solution varying from 1 to 7 N was converted to americium dioxide. Americium oxalate was precipitated from 0.1 N HCI with 100% excess oxalic acid and was converted to the dioxide by calcination at 800 ts C in air. The solubility losses in the oxalate precipitation filtrate averaged approximately 7 mg/liter of solution, with a total loss of 0.09%. (auth)

The mechanism of reduction for chromium(III) ion to chromium(II) at the mercury cathode was studied in 0.1M KNO/sub 3/. Data obtained at varying temperature and solution composition from polarograms gave values for DELTA H*, DELTA S*, and DELTA F* which indicated that two mechanisms were involved. At potentials more positive than the polarographic half-wave potential, the mechanism appeared to be simple electron transfer from the electrode to the chromium(III) ion in solution. When the potential was more negative than the half-wave, potential electron exchange between the reduced chromium ion near the electrode surface and a chromium(III) ion in solution became appreciable. Values for the heat of activation for the reduction of chromium(III) to chromium(II) in 0.1M KNO/sub 3/ for the electron transfer and exchange reaction mechanisms were determined to be 34.7 and 27.0 kcal mole/sup -1/, respectively. (auth)

The use of KEWB-B as a calibration source for pulse detection instruments is proposed. As a result, more information was needed: radial and axial distributions of fissions, number of fissions seen by the detection instruments, and materials seen by the detection instruments. The calibrations were achieved in six transient runs. (T.R.H.)