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Results of a conceptual design study for a curium powered thermoelectric generator of minimum size and weight which is capable of sustaining hard impact is presented. The generator produces a minimum of 13 watts of d-c power at 3 volts, and weighs 6.2 pounds excluding shielding. (J.R.D.)

The check valve for the Hallam Power Reactor uses a knife-edge bearing for the flapper in place of the usual journal-type bearing. Mechanical cycling in sodium at 600 deg F was used to check operation of this bearing. A total of 309 mechanical cycles was completed with no apparent malfunctioning of the valve. Measured leskage rates were 0.46 gpm at 0.93 psig, 0.73 gpm at 3.4 psig. and 0.32 gpm at 5.9 psig. (M.C.G.)

The HTGR prototype plant (Peach Bottom Power Reactor) is being designed to produce steam at l450 psi and 1000 deg F and to have a net capacity of 40 Mw(e). The fuel temperatures and gas pressures will be approximately the same as those required for larger plants. The reactor data and operating conditions for the graphite-clad core are given. The reactor and primary coolant systems are described. The prospects for development of the graphite-clad fuel element in time for use in the first loading of the reactor were improved by important advances in methods of fabrication and testing of both fuel compacts and graphite sleeves. The hot-pressing process for making fuel compacts was used successfully to make full-size compacts with a uniform distribution of ThC/sub 2/- UC/sub 2/ particles. Three irradiation capsules were fabricated and inserted in a test reactor to determine fuel compact and sleeve performance under HTGR conditions of irradiation and temperature. Two of these ran satisfactorily for the scheduled time of operation. A scope design study of the in-pile loop that will be used to evaluate the full-diameter graphite-clad element was completed. Experiments to determine the extent of fuel migration within the element were undertaken. Preliminary results …

For presentation at ASCE Convention in Reno, Nevada on Thursday, June 23, 1860. A description of the Hallam Nuclear Power Facslity is presented. The history of the project, program participants, site description, component development program, reaetor building, reactor structure, reactor core, sodium systems, instrumentation and control, fuel and component handling, auxsilary sustems, special design features, and advantages of sodium graphite reactor systems are discussed. (M.C.G.)

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.

A thermoelectric generator which produces 100 watts of electrical power continuously over a six-month operational life in a space environment was designed. It employs the heat produced by the decay of Cm/sup 24/ as the source of power. Uniform output over the operational life of the generator is accomplished by means of a thermally actuated shutter which maintains the hot junction temperature of the thermoelectric conventer at a constunt figure by varying the amount of surplus heat which is radiated directly to space from the heat source. The isotopic heat source is designed to safely contain the Cm/sup 242/ under conditions of launch pad abont and rocket failure, but to burn up upon re-entry to the earth's atmosphere from orbital velocity. (W.L.H.)

EWR-ESE-144, Fifty MC-890 type final assemblies were subjected to irradiation at low temperature test in order to determine if irradiation would reduce the frequency of high voltage breakdowns. A control lot of a like number of units was subjected to low temperature test in the same manner except without irradiation.

ESE-171: This EN run was made to determine if there is any relationship between the H-5 transistor electrical characteristics, the assembly processes, and slow warm-up of the MC-890 Power Supply. Included is the test data on the H-5 transistors and the results of the High Temperature Life Test (185 {degrees}F) of 380 MC-890 Power Supplies.

EWR No. VTE-188--Tubes processed by flashing getters immediately prior to seal-off from vacuum systems are compared for total residual gas pressure to tubes processed by flashing getters after tubes were sealed off vacuum systems. Comparisons of residual pressures determined from current flows in the cold cathode ion gauge.

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ABRAC is a three dimensional nuclear thermal depletion program to study the effects of water moderator density changes, resulting from flow variations and boiling, on neutron flux distribution and depletion. The program requires an IBM-704 with a memory of 32.768 words and ten tape units. (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 …

A micrometer that utilizes eddy current techniques is described. The gage is capable of measuring nominal 0.5000-in. aluminum rods to an accuracy of plus or minus 0.00005 in., and is unaffected by residual nonconductive surface filins such as oxides or corrosion products. (auth)

The reduction characteristics of plutonium trichloride are investigated. A flowsheet for batch reduction with Ca is included. (J.R.D.)

The concentrations of and activities due to N/sup 16/ and N/sup 17/ in the HFIR primary coolant water were calculated. At the pressure vessel exit, the N/sup 16/ activity is 3.9 x 10/sup 6/ dis/secml and the N/sup 17/ activity is 6.9 x 10/sup 2/ dis/sec-ml. Comparison of the N/sup 16/ activity with the data obtained from the ORR water system indicates that the calculated results are slightly conservative. (auth)

An estimate was made of the fission product activity which would result in the HFIR primary coolant system following a meltdown of the fuel element within the reactor. The rare gases and the halogens appear to be the main contributors to the gamma activity in the coolant system imnmediately after the meltdown, and iodine appears to be the main contributor 24 hours after the meltdown. (auth)

The 9 single-crystal elastic moduli pertaining to principal crystallographic axes of alpha U at 25 deg C were determined from measurements of high-frequency wave velocities for 21 modes in seven single-crystal specimens, using the phase-comparison method of McSkimin. From the results the elastic compliances, compressibilities, and Poisson"s ratios were computed for the principal axes. Th variations with crystal direction of the stiffness moduli, Young's moduli, and rigidity moduli were plotted. The nature of the anisotropy for the different moduli indicated that the nearest neighbor interatomic bonds are considerably stiffer than the next nearest bonds, which are only slightly larger in interatomic distance. (auth)

The admittance and transfer functions of large, solidcore electromagnets were determined. The effects of eddy currents and hysteresis were considered in deriving the functions. The study was concerned with the type of magnet employed in nuclear physics research which requires very precise regulation of the magnetic field. The study originated during the design of an analyzing magnet regulator for the ORNL 63-inch cyclotron. (W.D.M.)

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Preliminary design calculations were performed for a once-through type steam generator and reheater for advanced sodium power plants in the 300-Mwe range. Parameters and performance data are presented. (D.L.C.)

An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of …

The results of a study are presented which provided additional experimental-loop irradiation space for the AECDRD testing program. It was a premise that the experiments allocated to this reactor were those which could not be accommodated in the MTR, ETR, or in existing commercial test reactors. To accomplish the design objectives called for a reactor producing perturbed neutron fluxes exceeding 1O/sup 15/ thermal n/cm/sup 2/-sec and 1.5 x 1O/sup 15/ epithermal n/cm/sup 2/-sec. To accommodate the experimental samples, the reactor fuel core is four feet long in the direction of experimental loops. This is twice the length of the MTR core and a third longer than the ETR core. The vertical arrangement of reactor and experiments permits the use of loops penetrating the top cap of the reactor vessel running straight and vertically through the reactor core. The design offers a high degree of accessibility of the exterior portions of the experiments and offers very convenient handling and discharge of experiments. Since the loops are to be integrated into the reactor design and the in-pile portions installed before reactor start-up, it is felt that many of the problems encountered in MTR and ETR experience will cease to exist. Installation of …

Development of a radioisotope-fueled thermoelectric power-conversion system for the SNAP program is discussed. The generator operates at somewhat less than 1% over-all efficiency, has a power output of approximately 1.45 watts, and has a specific power of 0.l42 watts per pound. The power-flattening device did not perform as expected. A second generator is also described which produces a maximum power of 1.6 watts, has a maximum efficiency of 1.20%, and has a specific power of 0.32 watts per pound (exclusive of the weight of the heat source). (auth)