C Reactor was shut down on a scheduled basis at 8:30 a.m. March 6, 1961 for the purpose of overboring 17 process channels. this report will cover that outage and discuss problems encountered in completing the tasks involved in overboring.

An investigation was conducted to determine which tubes of the 1A loop heat exchanger are leaking. Air pressure and probing tests are inconclusive and cannot be used to verify chemical sampling. (J.R.D.)

Descriptions are given of various procedures used in determining leaks in the tubes of the 1B heat exchanger. Air pressurization tests determined leakage and leak rate of nine tubes. The leak-location-detector-probe method was found promising for locating defects along the length of the tube. Results of the probalog, dye-penetrant, and ultrasonic tests proved inconclusive in determining leak locations. (B.O.G.)

When the construction of the confinement facilities for Hanford production reactors is completed, the potential for contaminating the environment with accidentally released fission products will be greatly reduced. This was demonstrated already when rupture debris burned in the rear face area of one of the reactors in which the fog spray was installed and ready for service. The fog spray effectively prevented the spread of contamination to the ventilation equipment downstream of the rear face enclosure. Also, the commercially- available absolute filters and charcoal beds will significantly decontaminate the exhaust air, stream of all fission products except the noble gases. Laboratory studies indicate that the filter system should remove at least 50 percent of all fission product halogen vapors and over 90 percent of the particulate matter released from a nuclear incident. In addition, the charcoal beds should provide an additional 90 to 95 percent decontamination of the halogens that manage to pass through the filter. It is not expected that any of the noble gases released will be removed by this system. Laboratory studies, in general, substantiate previous estimates on the release of fission products from overheated uranium fuels. The theoretical estimates are quite accurate for the volatile and …

This study was developed to provide a correlation for the evaluation of the pile C{sub RI}/Tube C{sub RI} ratio (alpha value) for each of the Hanford reactors.

The inherent stability of the EOCR at full power, exclusive of the heat exchanger, is investigated. Space and time dependent differential equations and transfer functions for the nuclear and thermal processes occurring in the reactor core and vessel are derived. Reactor stability is analyzed by application of the Nyquist stability criterion. The results show that the EOCR has an extremely large margin of inherent stability. (auth)

This report addresses the analytical chemistry of Cobalt.

An increase in the flow of water from the lower depths of the Grand Coulee Reservoir was used to lower the river temperature at Hanford Atomic Products Operation (HAPO). A net average daily reduction of over 1 C resulted. The average for one day varied from 1.7 to 0.2 C. Before the dam control period, the Bonneville Power Administration transferred load from Grand Coulee Dam to other dams in order to conserve cold water. The authors calculate that there was more value to HAPO from this transfer than the combined effects of river bed losses and/or any variance in scheduled or base loads.

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Boron carbide-graphite nuclear control-rod materials containing up to 45 wt% B were prepared by hot-pressing techniques. The densities of these compacts decreased with increasing B content. Compact strengths, however, were noticeably increased with B content of 20 wt%, suggesting that the B acts as a sintering aid. Thermal stability was evaluated by annealing compacts under a variety of temperature and atmospheric conditions. The upper limit of the ioss of 13 was 5% to 10%~ in 200 hr at 2000 deg C--considerably less than would be predicted from the Langmuir evaporation equation. Diffusion of B atoms through the pores of the graphite, rather than evaporation from the surface, was found to be the rate- controlling step in the loss of B. Compacts containing 20, 30, and 40 wt% B were irradiated at 350 and 700 deg C to 1.0 and 2 4 x 10/sup 21/ nvt(fast). Physical integrity of the compacts was good after irradiation. Slight increases in tensile strength were noted in several instances. (auth)

In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

Carbon-14 is by far the most popular of the labeled isotopes because of its adaptability to so many chemical compounds and its relatively harmless characteristic beta radiation. Because the inert gas atmosphere contains up to 30% nitrogen, carbon-14 is generated within the K Reactors by the conversion of nitrogen-14, and if not immediately oxidized, carbon-14 may remain in the reactor core for some time. It is eventually oxidized and escapes when the outward gas leakage from the reactor is vented to the stack. Production Test 124 was authorized to determine if carbon-14 could be extracted from the K gas as a profitable by-product. It was concluded that as much as 500--600 curies per year of carbon-14 could be recovered if both K Reactors were provided with relatively simple extraction facilities. This is about twice the amount of this product for which there is a current market at the present price level of approximately $3,000/curie. The grade (as specific activity) of the product recovered in the production test was low -- ranging from 3 to 26 millicuries per mol. Substantial upgrading (perhaps by a factor of 10X to 100X) is possible by skimming techniques and reduced oxidation of normal carbon at …

A CDC-1604 subroutine package was written to facilitate the plotting of curves and points on linear, logarithmic, and semilogarithmic graphs using the CALCOMP plotter. The subroutines accomplish the necessary computations and prepare a magnetic tape for use by the plotter. (auth)

Aqueous zirconium fuel processing studies, directed at adapting the hydrofluoric acid process to continuous dissolution-complexing in order to increase the capacity of the ICPP process, resulted in determination of dissolution rates over a wide range of conditions. The addition of oxidants to 4.8M hydrofluoric acid used for the continuous dissolution of 3% uranium-Zircaloy- 2 fuel decreased the amount of uranium-containing residue in the dissolver at the expense of a decreased dissolution rate. Oxidized fuel was dissolved in 4.8M hydrofluoric acid by use of heat to initiate the reaction: the oxide film remained behind as a dissolver residue. Uranium in the dissolver product was oxidized to the extractable form by one minute treatment at 92 deg C with aluminum nitrate complexer solution: only 59% was extractable after 24-hour treatment at room temperature. Several hundred hours operation of a continuous bench scale unit indicated that uranium buildup should not be a problem in the dissolver since an apparent equilibrium was soon established. After 650 hours of operation, a Monel dissolver showed some signs of localized corrosion but exhibited generally good resistance. Preliminary results in a study of aluminum alloy dissolution rates, initiated because of unusually slow dissolution of certain aluminum-based fuel elements, …

To assure that the NEERVA reactor will function reliably during its test program, cleanliness must be controlled at each stage. These stages include design, manufacture, assembly, shipping, and test. The assembly stage is that phase which this handbook is concerned.

This report addresses command links.

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.

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Tests were conducted to compare the specific isotopic fission product activities with those of delayed neutron emitters which are monitored by the FEDAL System, and to obtain irformation on the source of activity observed in the PWR coolant. The monitors detected definite activity bursts of delayed neutrons during load increases, which indicated a rupture of the core blanket element. A radiochemical analysis of the coolant data did not show any unusual activity bursts, thus indicating a fuel element failure. It was concluded that the discrepancy in measurements is caused by methods not specifying exactly the time for taking radiochemical samples. The data are tabulated as activity and power output, and given as a function of time. A comparison is given of the activities in the primary coolant for April, June, and July of 1958. (B.O.G.)

This report explains a mathematical algorithm designed to complete contour descriptions of weapon components which have been only partially specified by free-form data.

Data transmission over telephone lines involves the acceptance of serialized binary data, modulation of the data to a signal suitable to the bandwidth of the telephone. channel, transmission of the signal over the line, reception at the receiving end and demodulation of the signal to reproduce the original binary data. One typical application of data communication is between a low-speed terminal and a computer. In such a case, full-duplex operation is often desirable (i,e., transmission of information on the line in both directions simultaneously). The data set described in this report operates full-duplex and provides for channelization of the telephone line for the simultaneous use of several data sets. The data set can either be coupled to leased telephone lines or acoustically coupled to any telephone.

The authors calculate the branching ratio {Lambda}(P {yields} {ell}{sup +}{ell}{sup -})/{Lambda}(P {yields} {gamma}{gamma}) to lowest contributing order in quantum electrodynamics, with a vector meson model for the pseudoscalar meson form factor. They treat the processes {eta} {yields} {mu}{sup +}{mu}{sup -}; {eta} {yields} e{sup +}e{sup -}; K{sub 2}{sup 0} {yields} {mu}{sup +}{mu}{sup -}; K{sub 2}{sup 0} {yields} e{sup +}e{sup -}; {pi}{sup 0} {yields} e{sup +}e{sup -}. Results are compared with those of previous calculations.

This proposal is submitted in response to Request for Proposal No. 438468, dated 27 May 1969, from the Power Procurement Section of the NASA Lewis Research Center, to design, fabricate and test Primary NaK Loop (PNL) valves for use in ground testing of the SNAP-8 Power Conversion System.

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)