Despite the fact the adequate data from long-term, chronic experiments on the effects of Po{sup 210} are not available, some sort of a working figure for the maximum permissible body content of polonium is desirable. As described herein, calculations involved in determining the permissible body content of polonium generally fall into three classes. 1. Comparison of X-ray and polonium toxicity and application of an acceptable X-ray tolerance figure. 2. Assumption of a most sensitive organ and computation of the critical concentration of radioactive material present in this organ under conditions such that the organ received an assumed maximum permissible exposure over an indefinitely long period of time. 3. Comparison of radium and polonium toxicity with application of an acceptable maximal content of radium.

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1951. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

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Heat transfer coefficients were determined for the eutectic mixture LiF- KF-NaF (Flinak) flowing in forced convection through circular tubes. Heat, electrically generated in the tube wall was transferred uniformly to the fluid during passage through small-diameter tubes of nickel, Inconel, and 316 stainless steel. The variables involved: Reynolds modulus (N/sub R//sub e/), 2300 to 9500; Prandtl modulus (N/sub P//sub r/, 1.6 to 4.0; average fluid temperatures, 980 to 1370 deg F; and heat flux, 9,000 to 192,000 Btu/hr-ft/sup 2/. Forced-convection heat transfer with Flinak can be represented by the general correlation for heat transfer with ordinary fluids (0.5 < N /sub P//sub r/< 100). The existence of an interfacial resistance in Flinak-Inconel systems was established and its composition determined. Preliminary measurements of thermal conductivity and thickness of film were made. The results verify the effect of the film on Flinak heat transfer in small-diameter Inconel tubes. Thermal entry lengths, determined from variations of local heat transfer coefficients in the entrance of the heated section, were correlated with the Peclet modulus. (auth)

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Methods of melting uranium and its alloys are examined including vacuum induction melting and consumable and non-consumable electrode arc melting. The equipment and procedures are outlined with the limitations and uses of each method. In addition, the preparation and properties of Al--U, Cr-U, Mo--U, U-- Nb, U--Si, and U-Zr, and Nb--U--Zr alloys are described. Special techniques of agitation during freezing to produce fine castings are listed along with a discussion of the advantages of centrifugal casting. (J.R.D.)

The objective of this supplement is to authorize a change in the panellit trip range from 25--75 psi to 5--95 psi. The test hole facility consists of two concentric aluminum tubes which extend from the front face to the rear face of the reactor. The ID of the inner tube is 2--7/8 inch. Water from one crossheader supplies the annulus, water from another crossheader supplies the inner tube. The three-foot-long, .570 inch OD fuel element is centered in a 40-inch long aluminum holder which has an ID of 1.380 inch and an OD of 2.800 inch. The panellit gage which monitors the flow to the inner tube fluctuates to such an extent during start-up that on two occasions the reactor was scrammed. During equilibrium operation the panellit gage reading remains stable. A possible explanation of this behavior is that during start-up aluminum spacers which are in the inner tube as part of the test charge chatter and cause variations in the water path through the tube. It is further surmised that at equilibrium operation the pressure drop across the column in the tube is sufficient to suppress the chattering. It is concluded that extending the trip range to 5--95 psi …

In the fabrication of the steam generator on APPR-1A it was considered necessary to roll the Type 430M tubes into carbon steel tubesheets to ASTM Specification A350-Grade LF-1, modifled with 1.66% nickel; and weld the tube ends to the stainless steel overlay previously applied to the tubesheet. The rolled joint was a necessary precaution to prevent secondary water, that might contain chlorides, from contacting the stainless steel weld joining the tubes to the tubesheets. The welded joint provided the mechanical strength for attaching the tubes to the tubesheets. A laboratory program was conducted, therefore, to develop practicable procedures for welding the Type 430M tubes to the stainless steel overlay; as well as to assure that the tubes could satisfactorily be rolled to the tubesheets. Automatic and manual tungstenare welding procedures were developed that were capable of consistently providing an austenitic weld having a strength exceeding that of the heat affected zone or the unaffected tube itself. Type 430M tubes in the asreceived, and softened conditions were rolled into prototype test units under various conditions of rolling. It was concluded that the Type 450M tubes in the as-received condition could be satisfactorily rolled into the A360Grade LF-1 tubesheet and be tlght …

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.

K area emergency water backup studies have been hampered by poor data on flow through the reactor under various emergency conditions. Various tests have been run where emergency conditions have been simulated and flow measurements attempted. In all previous tests, the accuracy of the flow measurements have been questionable. Flow from the high-pressure crosstie can be measured by an orifice in the crosstie, but there has not been any method of measuring the service water contribution to total reactor flow under simulated emergency conditions. One method of measuring the total reactor flow regardless of its source is to determine the relationship between total flow through the reactor and the bottom of riser pressure. After this relationship has been determined for the flow range of interest, then flow to the reactor can be determined by reading bottom of riser pressure (BORP) and converting that to flow. The objective of this production test was to obtain the relationship between BORP and total reactor flow in the range of 10,000 gpm to 25,000 gpm. An additional objective of this test was to check the accuracy of the No. 2 pump discharge venturi.

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Earlier this year we briefly discussed the potential hazard of incurring an inadvertent uranium fuel element fire during discharge. This letter will provide data which will be of assistance to you in assessing the potential hazard, and in establishing charge-discharge procedures to minimize the probability of an irradiated fuel element lodged in the discharge area reaching aluminum jacket melting temperature without detection.

A nonnuclear explosion involving an evaporator occurred in a shielded cell in the Radiochemical Processing Pilot Plant at Oak Ridge National Laboratory on Nov. 20, 1959. Plutonium was released from the processing cell, probably as an aerosol of fine particles of plutonium oxide. It is probable that this evaporator system had accumulated -1100 g of nitric acid-insoluble plutonium in the steam stripper packing; the explosion released an estimated 150 g inside Cell 6, with about 135 g in the evaporator subcell, and about 15 g in the larger main cell. No radioactive material was released from the ventilation stacks; no contamination of grounds and facilities occurred outside of a relatively small area of OaK Ridge National Laboratory immediately adjacent to the explosion. No one was injured by the explosion, and no one received more than 2% of a lifetime body burden of plutonium or an overexposure to sources of ionizing radiation either at the time of the incident or daring subsequent cleanup operations. The explosion is considerdd to be the result of rapid reaction of nitrated organic compounds formed by the inadvertent nitration of about 14 liters of a proprietary decontaminating reagent. In cleanup the contamination was bonded to the …

In coolant system such as NPR's, the coolant activity level increase with each succeeding pass through the reactor flux until a saturation limit is reached. Therefore, the activity level of the NPR coolant system will be much higher than that of the old reactor once-through systems. This report is the determination of the specific activities (disintegrations/cc{center dot}sec) of the various coolant impurities which determine the total activity of the coolant system. 10 refs., 13 figs., 2 tabs.

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This report provides technical specifications applicable to the eight operating production reactor facilities, namely B, C, D, DR, F, H, KE and KW (referred to as the Hanford Production Reactors), located at the Hanford Site in the State of Washington. Basic descriptions of the reactor facilities are also contained in the report.

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Pressure drop tests were performed in the 189-D-Hydraulics Laboratory to determine the number of K-Reactor solid aluminum pieces (SAs) required to throttle the tube flow of a 46-piece KVE and a 38-piece KVN arch-rail I&E fuel charge to that of a corresponding bridge-rail charge. In addition, the effects on Panellit pressure and tube flow were determined for several cases.

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

This report presents details of the activities of Douglas United Nuclear at the Hanford site during the month of January 1968.

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A Reed Tool Compsny 73JA drill bit used in the Geysers geothermal field failed primarily because of severe frictional wear of the nose bearing. There is no indication that well temperature by itself caused failure of any part of the bit. However, the high well temperature together with the friction heat generated produced a temperature which lead to softening and premature wear failure of the nose bearing.

Progress is reported on research projects in the fields of radiation protection, industrial hygiene, instrument development, fire safety, decontamination, and environmental protection. (HLW)