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Production of Pu nitrate, UO{sub 3}, and unfabricated Pu metal met schedules. Decontamination performance of Purex process continued below standard. The cerium-144 cask is being redesigned. A ``powered ferret``, for driving a scintillation counter through a conduit to monitor ground activity beneath waste storage tanks, is being designed.

R and D is reported in the following: Reactor and Fuels (PRTR, Pu fabrication pilot plant, KER, NPR, materials); Chemical R and D (Pm recovery, fission products, Purex column, non-production fuels reprocessing, Salt Cycle process); Physics and Instrument R and D (PCTR, NPR, critical experiments, PRTR); and Biology (monitoring, irradiation experiments).

This document details activities of the Fuels Preparation Department during the month of January 1960. (FI)

This document details activities of the irradiation processing department during the month of January 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.

The Criteria for Speed of Control must be met for all cases being considered in the Expansion Studies. These Criteria are dependent upon the ratio Volume of water/volume of uranium in the process channel, the reactor power level, and the rate at which the vertical rods are released and inserted into the reactor. For those cases wherein the graphite channel is overbored by 0.200 inch., the criteria can generally be met for the power levels under consideration since the ratio of Vw/Vu is maintained adequately small by the use of large diameter fuel elements. If the graphite is not overbored, high pressure drop across the fuel element section must be maintained in order to maintain small values of Vw/Vu, which results in a substantial increase in the front header pressure. A method is developed in this document which ties together and defines the hydraulic and physics characteristics of the fuel-process channel geometry necessary to satisfy the Speed of Control Criteria over a range of feasible operating conditions. This study assumes no graphite overboring and employs throughout a smooth bore zircaloy process tube having an inside diameter of 1.650 inches. The fuel elements are self-supported., and the inner hole size was …

This monthly document details activities of the Hanford Operations Office during the month of January 1960. (FI)

Nickel plating on the coolant contacting surfaces of aluminum-jacketed fuel elements is highly attractive for increasing resistance. Potential benefits include a highly corrosion-resistant coating for severe localized conditions, reduction of mechanical damage to fuel element jackets, improved fuel element alignment (by reducing friction between fuel element and process tube ribs) and probably lower overfall surface temperatures to reduction in corrosion product film with improved corrosion resistance, neutron economy might also be realized. For example, substitution of a 0.5 mil thick nickel plate for 15-mils thickness of aluminum jacket would result in no reactivity loss and permit a concomitant increase in uranium volume, or in coolant flow annulus. Attendant problems include providing an adherent continuous plate of uniform thickness and possibly contamination of reactor effluent by radio-nickel-cobalt, and phosphorous and it was found that gross sloughing of the nickel plate had occurred. Development and testing work was carried out to determine the cause and a solution to the Greece problem. Studies were limited to the behavior of chemically-deposited nickel because of the unique capability of the process to deposit a coating of uniform thickness in the 0.1 - 0.2 mils thick range, regardless of the geometry of the plated piece. Based …

This report presents the results, conclusions and recommendations of a comprehensive study of the hazards potential of the 1706-KER in-pile recirculation loops, Although more data are needed to verify or refute certain tentative conclusions, the general conclusion of this report is: with the exception of severe coolant loop rupture incidents, especially process tube ruptures, the loops will not possess serious hazards potential when they are operated within the limits specified in the Process Standards as amended by the acceptance of the recommendations of this report concerning the pressurizer vent system and the process tube exit trips. The potential for fuel burnout or meltdown following a coolant loop rupture should be reduced considerably after completion of Project CGI-839, ``Modification to Fuel Element Test Facilities.`` The present potentially ``worst case`` coolant loop rupture could lead to an accident comparable in severity to a single K-Reactor tube rupture and meltdown. Serious personnel and property contamination would be confined to the vicinity of the reactor following completion of Project CGI-791, ``Reactor Confinement`` unless unusual atmospheric conditions persisted during the entire release time, i.e., strong, unidirectional winds causing rapid straight-line, non-mixing travel of the fission products plume. However, graphite stack damage from a tube rupture …

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.

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.

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.

Cavitational flow has been known to exist in-rear crossheader ``Parker`` fittings at B, D, and F Reactors for the last five or six years. Calculations showing initiation of cavitational flow as a result of high flow rates in the present fittings were verified by experimental data in 1954. A study is currently being conducted to determine the required plant modifications to obtain flow increases on the order of fifty percent above existing flows. This study and the results of preliminary tests that show nominal flow increases may be obtained by reaming rear crossheader fittings has focused attention on the condition of existing rear face piping. To obtain an estimate of the effect of cavitation in rear crossheader fittings resulting from past and current operating conditions, twenty one fittings were examined during the period October 6, 1959 to November 30, 1959. This document reports the inspection results.

Cobalt samples frequently are irradiated in nuclear reactors to produce gamma sources and can be irradiated as integral flux monitors because of the long half-life of the isotope produced. At the present time a small cobalt sample is being irradiated within the KW Reactor Snout facility for future use as a radiographic source for inspection of finished product in the Chemical Processing Department. Analysis was made to estimate the buildup of activity in this sample; the general equation may be of interest and value for other cobalt sample irradiations.

The average energy release per fission event in a reactor is dependent on the composition and arrangement of the lattice materials. In a study of heat generation in the NPR, Nilson developed expressions for calculating the average energy released in each material per fission event. These relationships have been used in the present calculations to obtain the energy release per fission in existing Hanford reactors.

A significant reduction in non-equilibrium losses is possible through the use of poison splines for reactivity and heat distribution control during reactor startups. The curves presented show the results of an analysis of recent KE Reactor poison spline startup usage. These curves demonstrate the magnitude of gains possible at other reactors through the use of poison splines for turnaround control.

A study was made to determine the optimum arrangement of the fringe flow zones at KE Reactor. Also considered, was the possibility of converting part or all of the existing fringe zone (actually three flow zones) from solid metal to I & E metal to decrease rupture potential in these low flow zones. The necessity for this study was indicated by high tube outlet temperatures in the fringe flow zones and the recent occurrence of two solid metal ruptures in process tubes located in the fringe zone of the reactor. Two additional solid ruptures occurred during the period that conversion to I & E in zone 2, as recommended below, was being completed.

Steady state hydraulic demand curves were obtained for tube powers of 500, 1000, 1500 and 2000 KW with an inlet water temperature of 20C and a rear header pressure of 25 psig. These curves are shown in figures. The point of initial unstable flow for various tube powers is shown for a front header pressure of 325 psig. The flow rate that would lead to the initial point of unstable flow as a result of a sudden plug upstream of the Panellit tap is shown in a figure.

Canning bath Al-Si, supplied from offsite vendors and reclaimed lathe turnings in the 313 building, is used in the production of I & E fuel elements. A study was made of the effect of increasing the Ni content to over 0.5% in the canning baths, in order that all of the X-8001 scrap could be reclaimed. Effect on bond quality, weld integrity, and canning bath operation was studied. Based on adverse weld quality, slight loss in reactivity, and potential for furnace channel plugging, it is recommended that the present Ni specification of 0.5% maximum remain unchanged.

A new rear nozzle has been developed by Mechanical Development (A), IPD to alleviate the stuck gun barrel problem on the process tubes at F reactor. This nozzle is called the 105-F rear face floating nozzle (RFFN) and is shown in Figure 4. The purpose of this study was to determine the hydraulic operating characteristics of this nozzle and to compare these characteristics with those of the standard nozzle assemblies. This report is essentially an addendum to EW-63756 1 which presented energy loss data for various outlet fitting assemblies for possible use on the BDF type reactors The experimental data for this study were obtained with the use of the hydraulics mockup in the 189-D Thermal Hydraulics Laboratory.

This report examines the pertinent features in the hazards analyses of the Hanford Reactors which may be affected by the substitution of zirconium tubes for the present aluminum process tubes. Resized I & E slugs, designed to preserve present pressure drops across the active zones and to minimize corrosion, are used as examples to compare the characteristics of the zirconium tubes reactor with the present.

This report is essentially an addendum, to an earlier report which presented energy loss data for various enlarged outlet fitting assemblies for possible use on the old reactors. The current report presents data for another candidate outlet fitting assembly which would tend to reduce the energy losses in the process tube outlet fittings. This outlet assembly was developed by the Materials Development Operation, IPD. The data show that the flexible connector assembly would result in a 6.9 per cent increase in normal single tube flows without modifications to existing headers or nozzles or a 13.5 per cent increase if the existing header (Parker) fittings were reamed to 0.610 inch ID. These percentages are based on comparison with the standard helical connector outlet assembly.

Experiments have been conducted in the Hydraulics Laboratory, at the request of IPD`s Mechanical Development-A Operation, to determine the energy losses of various enlarged outlet fitting combinations. These experiments were conducted an steady state runs and allow the determination of the normal operating point (flow rate) of a reactor process channel under selected conditions of front header pressure and fuel charge. No attempt is made to make a mechanical or economic evaluation of the particular fitting combinations, although observations were noted which might bear on this evaluation. It is very important for the reader to bear in mind that changing outlet fittings will definitely affect the reactor tube power limits and outlet vater temperature limits. The size of the outlet fittings largely determines the present outlet temperature limits of the old reactors. The flow characteristics of these present fittings cause some degree of pressurization to suppress boiling on the fuel charge and also cause dual Panellit trip protection for certain flow changes and for power surges. Enlargement of the outlet fittings may actually reduce the allowable outlet coolant temperature limits. Since these effects cannot be determined on the apparatus used in these experiments, a complete discussion of this point is …

The thermal and total epithermal neutron flux distributions in the EGCR from the center of the core through the biological shield were calculated. The maximum values of the flux distribution are presented in graph form. Ordinary concrete was found to permit a thermal flux buildup, similar to that found in the graphite reflector. In heavy concrete there was no such buildup seen, since the thermal flux is attenuated through the complete shield. A Val Prod code using multigroup diffusion theory was used to calculate the neutron flux distributions. Calculations are shown. (M.C.G.)