A class of linear proportional servomechanisms is examined in which an electrical signal proportional to output force is used to improve performance. The effect of this "force feedback" on a positional servomechanism is analyzed as well as the effect on a special type of servomechanism which reflects load forces back to the input. This latter type of servomechanism is called "force reflecting." Laboratory models of these servormechanisms were designed and constructed, and experimental data are presented in support of the analysis. (auth)

A study of the dynamic behavior of the Merchant Ship Reactor Pressurized Water Loop was made using the Reactor Controls Analog Facility. Computer curves show the predicted response of the loop temperatures to normal load changes and component failure accidents. Except for complete flow stoppage, which was not investigated here, the safety system was shown to be adequate in curbing loop temperature excursions due to postulated accidents. (auth)

Tests were performed to determine the effect of thermal cycling on the across-the-seat leakage characteristics of valves for use in the sodium coolant system of the Hallam Nuclear Power Facility. Four 12-in. valves were cycled between 680 and 1100 deg F in a sodium loop and periodically tested for across - the-seat sodium leakage. The valves tested were two throttling (ball) valves and two blocking (gate) valves. The first-ball valve had a consistent leakage pattern which vas primarily a function of valve temperature, but also varied with pressure differential and number of thermal cycles. The second ball valve showed high initial leakage rates which became excessive after some sodium circulation. Using data obtained with the first-ball valve, a general equation for the rate of leakage across the valve seat was obtained for this type and size of ball-valve as a function of temperature, pressure differential, and number of thermal cycles. No sodium leakage was exhibited by either gate valve throughout the test. None of the walves exhibited any leakage through the freeze-seal along the stem. (auth)

A method is described for observing the uptake of Fe by rat bone marrow cells in vitro. Results of experiments on effects of anemic serum and human erythropoietin are presented with a brief discussion. It is concluded that the differences in uptake of Fe/sup 59/ are the result of isotope dilution. (auth)

Because of the danger of precipitation of polymerized Pu/sup 4+/, a research program was initiated for studying the polymerization of Pu/sup 4+/ in the UO/sub 2/(NO/sub 3/)/sub 2/--HNO/sub 3/ system. Th e present state of knowledge concerning the polymerization of Pu/sup 4+/ is reviewed, and the program is discussed in detail, including equipment and procedures. The program is evaluated from the viewpoint of hazards, health physics procedures, and personnel safety. (D.L.C.)

Fission prcduct cross sections were measured radiochemically and mass- spectrometrically for gold bombsrded with 112-Mev C/sup 12/ ions. Cross sections for 43 nuclides were measured for elements from nickel to barium. Thirty-six yields are either primary fission product yields (independent yields) or were corrected (with less than 25% correction) so as to represent independent yields. The independent yields were empirically systematized and a yield-mass curve was constructed. The yield-mass curve is compared with the yield-mass curves obtained from the fission of Bi with 22 and 190-Mev deuterons. The yield systematics indicate that the sum of the mass numbers of complementary fission products is 13 plus or minus 1 amu less tban that of the compound nucleus, and the sum of the charges of complemertary fission products is two units less than that of the compound nucleus. It is postulated that 9 plus or minus 1 neutrons and an alpha particle must have been emitted. Evidence is presented that at least three and possibly more of the neutrons are emitted prior to fission. The most probable charge of the fission products as a function of mass number was determined empirically. It is shown that from mass number 80 to mass number …

The chemical vapor deposition of niobium, molybdenum, tungsten, chnomium, carbcn, and niobium--vanadium alloys in a fluidized bed of UO/sub 2/ powder particles wss used to provide uniform, dense, nonporous coatings on the individual particles. in the case of niobium, which received major attention, hydrogen reduction of niobium pentachloride vapor was used as the vapor- deposition reaction. The most serious problem was that of maintaining bed fluidity aad avoiding agglomeration. This problem was overcome to permit routine operation of the coating equipment. In the entire program of 68 experimental runs, only 1.1 per cent of the product was lost by agglomeration. In routine operation, this loss should be even lower. (auth)

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The neutrons in an operating nuclear reactor are generated primarily by the fission events which are taking place. The great bulk of these relatively high energy or fast neutrons are slowed down or thermalized by a series of elastic collisions with the moderator nuclei which comprise the bulk of the volume of the reactor core. Once slowed down, the neutrons diffuse through the reactor core until they are absorbed or eliminated by some other process. Each of these three groups of neutrons, i.e., the fast or source neutrons, the intermediate or slowing down neutrons, and the slow or thermal neutrons, has a characteristic energy distribution. At a constant power level or rate of fissioning, an equilibrium is soon established among the groups at any point in the reactor. If it is assumed that a smooth transition exists between the different energy groups, it is possible to evaluate the entire neutron spectrum at a point in the reactor by determining the parameters which characterize each of the three groups. This has been done in the F Reactor Quickie Facility using radioactivants.

Radioassays of milk, soil, grass, and other material located near the Savannah River Plant showed no detectable Sr/sup 90/ due to plant wastes. These assays indicated that the Sr/sup 90/ found in this geographical area (25-mile radius of SRP) was nuclear weapons debris. Sr/sup 90/ was separated from milk by absorption on cation exchange resin. An extraction technique was used in assaying Sr/sup 90/ in all types of samples. Cs/sup 137//Sr/sup 90/ ratios were found to be constant in soil samples. (auth)

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The high pressure loop installed in the 3X3 reflector position of the ETR and the associated instrumentation to detect and study failure mechanisms handled the rupture tests without difficulty. Failure of the elements was initiated by shearing off a projection on the fuel elements. The first test of the series used previously unirradiated seven-rod clusters. After the projection was sheared off the fuel elements were operated for seven hours with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second and third tests used fuel which had been irradiated to 2400 MWD/T at Hanford prior to insertion into the ETR. The second test was operated for 14 hours after the projection was sheared off--again with no failure. The third test was operated for only 33 minutes after the projection was sheared off before fission product activity in the loop water caused the test to be terminated.

The pericd from February, 1959, when CE assumed operative responsibility of the SL-1 Reactor Plant, to February, 1960 is covered. The operations of the year are summarized; the reactor, instrumentation, mechanical, electrical, and facility systems are evaluated; health and safety, and the operational costing program are discussed. (auth)

The Stationary Low Power Reactor No. 1 is a three Mw boiling water reactor designed to demonstrate the feasibility of a nuclear reactor to supply electrical power and space heat for remote sites. In addition to performance evaluation the facility provides training for military personnel. The reactor was operated for 1159 hr during the quarter for a total core burnup of 20.3%. Power generation was 103.7 Mwd for a total power accumulation of 466.9 Mwd. Eight malfunctions occurred during the quarter for a total unscheduled downtime of 40 hr 44 min, As a result of malfunctions, aluminum keys will be installed on the control rod drives, instrument well covers were removed, and an order has been placed for a station auxiliaries breaker with a higher temperature rating. Data were taken on seven tests during the quarter. Four of these are expected to be completed during the next quarter. All equipment items on order for the SL-1 power extrapolation expansion program are scheduled to be delivered before June 15, 1960. Although condenser dampers and damper controls, process instrumentation, equipment cabinets, and the radiation monitoring equipment have not yet been ordered, construction will not be delayed. The two tie-ins to the existing …

In the Hanford production reactors, failure of fuel element cladding is known as subcooled burnout since the bulk coolant temperature is below the saturation temperature at time of failure or burnout of the fuel cladding. The heat generation rate at which subcooled burnout occurs, establishes a limit to allowable reactor power level. This document compares present and potential power level limits at the Hanford production reactors with power level limits established by subcooled burnout.

Many of the experiments now being conducted on high-energy accelerators requires the use of beams of charged secondary particles. It is worth while at this time to attempt to summarize information about some of the most useful methods of setting up such beams. We are not concerned here with the primary beam of the accelerator. Rather, they assume that a target is struck by the primary beam and that it is desired to form a beam from the secondary charged particles that emerge from collisions within the target. The simplest system of forming this beam of secondary particles involves the use of magnetic fields only. In most cases it is desirable to obtain a beam of particles of known magnetic rigidity, or momentum. The bulk of this article is addressed to this problem. Some comments are also made about the use of electric fields in conjunction with magnetic fields. The inclusion of electric fields allows the separation of a beam of known momentum into its various components according to the velocities of the particles, hence according to the masses of the particles. These are referred to as ''separated beams''.

The numerical results of the IBM-704 computer can be put into graphieal form with the aid of the IBM-717 peripheral line printer, and programming techniques using FORTRAN II are described for the production of several kinds of graphs. Different symbols can be used, and a total of 100 horizontal spaces are available, giving a resolution of 1%. (D.L.C.)

The gamma rays emitted by ORR fuel elements and by the fuel section of shim rods are measured as a function of position along the elements with a very small graphite ionization chamber. Comparison of the fuel burnup calculated from the gamma measurements and by the flux-time method shows good agreement. This means that the gammaray distribution measurements could be a good method of determining the U/sup 235/ consumption in fuel elements. Distributions of the macroscopic absorption cross section and the infinite multiplication factor along fuel elements are computed from the gamma dose rate distribution. The limited usefulness of the shim rod fuel section is discussed in the last section. (auth)

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The rear face or discharge area of a reactor contains all the appurtenances necessary to discharge irradiated fuel, to collect hot coolant from each process tube, to monitor tube and effluent temperatures, and to monitor the coolant for ruptured fuel elements. Generally, failure of a rear face piping component would not affect the safety of the reactor since the coolant has fulfilled its purpose, that of cooling the fuel elements. The failure may, however, cause failure of one of the monitoring devices and if undetected could lead to a minor reactor incident. The Purpose of this report is to review all information generated during the past three years concerning the condition of rear face piping and hardware. This review includes the history of rear face piping and hardware problems, study activities taken to ascertain the condition of the components, action taken to correct actual component failures, programs recommended to correct deficiencies which operating experience and engineering judgement indicate are necessary, and programs to accumulate additional information to support design of new piping and hardware components.

Low attack rates (0.1 to 0.5 mpy) were displayed by coupon specimens of type 347 stainless steel, titanium RC-55, and Zircaloy-2 which were exposed for 2877.5 hr in an oxygenated slurry of Th--8% U oxide, 116.5 hr in water, 6.9 hr in 5% HNO/sub 3/, and 4.3 hr in 3% triscdium phosphate during mns SM-6 through SM-9 in the slurry blanket mockup. The leading coupon of type 347 stainless steel showed a slightly higher rate than the other stainless steel ccupons due to entrance effects. Specimens of SA-212-B carbon steel displayed average attack rates of 2.9 mpy. (auth)

An aluminum-nickel alloy-clad ceramic-fueled plate of the BORAX-IV type was examined destructively after irradiation to a maximum burn-up of 7.800 Mw days per ton in the circulation water loop in the MTR. Irradiation was at 600 psig and 465 deg F. with local boiling in the area of highest neutron flux and a maximum heat flax of 500.000 Btu/(hr) (ft/sub 2/). The specimen performed satisfactorily in spite of several factors which made conditions more severe than those expected to exist in a reactor fueled with this type of element. The corrosion rate of the aluminum-nickel cladding on the element was approximately 10 mils per year. Oniy slight breakup of the ceramic pellet fuel was experienced and there was no evidence that the degree of cracking was influenced to any appreciable extent by exposure to radiation. (auth)

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.