The diffusion of hydrogen in zirconium hydride was studied using permeation techniques. The rate of permeation of hydrogen through zirconium hydride disks was measured for small concentration gradients. Data were obtained at 61 to 65 at.% hydrogen and 500 to 750 ction prod- C. The diffusion coefficients were determined by the time-lag method. Ho variation of the diffusion coefficients with hydrogen concentration was observed. The diffusion coefficients can be expressed by D (cm/sup 2/ per sec) = 599 exp (-34,800/RT). (auth)

Production Test IP-272-A-FP authorizes the installation of up to 100 smooth bore Zircaloy-2 process tubes in C Reactor to demonstrate the feasibility of self-supported fuel elements for production use. An additional 200 zirconium tubes are expected to be delivered by mid-year and con be used to expand the initial demonstration facility. It is the purpose of this document to assess the status of the pilot demonstration program from the B-C Reactor Operation viewpoint.

The nuclear characteristics of the CETR are described. Core operating lifetime, control-rod worth, and powerdensity distribution are discussed in relation to maximizing the core operating life. Other objectives of nuclear design are to minimize the power-density variation and to assure control of the reactor. (J.R.D.)

On January 22, 1960, HW-63487 Piping Changes for Increased Production at B, D, DR, F, C, and R Reactors, was published. This study investigates the valve pit piping, front and rear face piping,and effluent lines between the 105 Buildings,and the retention basins to determine modifications necessary to increase power levels in the 100-B, D, DR, F, C, and H Areas by increasing either or both temperatures and flows. The study was based on detailed hydraulic and stress calculations of the existing and proposed piping systems which because of their detailed and voluminous nature were not included in the study. It is the purpose of this study to document, in so far as possible, these calculations together with pertinent information which was not included in the original study.

A developmental program was conducted to provide and in-pile loop facility for use in evaluating gas-cooledreactor fuel asubassemblies. The program included the design, construction, and installation of a recirculating gas loop which is located in a 6 by 6-in. facility in the aluminum reflector of the ETR. The loop system was designed to recirculate the primary nitrogen coolant at flow rates up to 0.9 lb per sec and pressures up to 200 psia. It will accept fuel subassmeblies up to 36 in. in length and 2.26 ia. in diameter with specimen power generation up to 150 kw. The maximum coolant temperature at the specimen outlet is set at 1500 deg F. The loop system includes the in-reactor section, the machinery, the control system, and the specimen-handling apparatus. Salient features of the re-ertrant system include an aluminum pressure wall in the in-reactor section, static gas insulation between the reactor coolant and the circulating loop gas, and a controllable rate of heat exchange between the specimen inlet- and specimen outlet-gas channels in sections of concentric countedlow piping. The three blowers in the system feature grease-lubricated bearings and water cooling. The complete system was tested out of pile and is now installed in …

Two types of aluminum-clad uranium alloy fuel elements, a square (parallel flat plate) and a circular (concentric cylindrical shell) were investigated to determine their relative suitability for use in the Piqua Reactor. Nuclear, thermal, and mechanical data are given, and considerations leading to selection of the circular element are presented. Design dimensions are listed and reactor thermal design and operating conditions are given for the proposed element. (auth)

This report discusses travel to the Oak Ridge National Laboratory on January 26 and 27 and the Savannah River Laboratory and Plant on January 28 and 29, 1960. The primary objectives of the trip were to obtain information on the containment philosophy for processing plants recently adopted by Oak Ridge National Laboratory and on the Palm Program at Savannah River.

Consolidated Edison type fuel pellets were irradiated and analyzed, to determine the extent of fracturing, particle size of fines produced, and the rate of dissolution in boiling 13M HNO/sub 3/-0.04M NaF -0.1M Al(NO/sub 3/)3. Fused sodium or potassium hydroxide was used to shatter the pellets at 400 deg C or higher. Similar pellets were dissolved in H/sub 3/PO/sub 4/ and fused ammonium bifluoride. An investigation was made of the thermodynamics and limits of flammability of gases expected during the dissolution of sodium-bonded stainlesssteel-clad fuels in aqua regia or sulfuric acid. The amount of hydrogen evolved during Darex dissolution of 304 stain less steel was studied as a function of the fraction of total dissolving time and the total gas evolved. The rate of dissolution of tin in HF containing H/sub 2/O/sub 2/ was> 10 mg/cm/sup 2/- min at 13 to 72 deg C, but decreased> 10 times at 13 deg C when HF was replaced by NH/sub 4/F. A technique was developed for disintegrating and leaching graphite fuels, which yielded a recovery of 99.85% + uranium from fuels containing approximately 5% uranium. The uranium extraction in the Immi hot- cell facility indicated a 0.33% loss in the mixer-settler using the …

Basic studies of the kinetics of the reduction of copper oxide were made to establish the effect of the solid phase on the over-all reaction kinetics The reaction CuO + H/sub 2/ at the only rea Cu + H/sub 2/O consisted of an induction stage, an acceleration or autocatalytic stage terminating at about 35% reduction of the exide, and a decreasing-rate stage The reduction rates for each stage were dependent on the nature of the initial oxide, the degree of subdivision of the oxide, and the temperature but were independent of the mass of the oxide phase. Addition of the reaction product copper had no measurable effect on the reaction. Water vapor in concentrations of 25 mg per liter of H prevented reduction at 112 ction prod- C The inhibiting effect decreased rapidly as the temperature was increased and disappeared entirely at 190 ction prod- C. Once reduction bad started. water vapor had practically no effect The acceleration and decay stages were very closely approximated by a semiempirical equation based on the initial reaction occurring on certain active nuclei followed by a rapid growth of these nuclei by a branching-chain mechanism. The reduction rate reached a maximum and subsequentlv decreased …

A lecture intended to provide a general background in weapons effects is presented. Specific areas of nuclear explosion phenomena pertinent to the design of hardened systems discussed include nuclear radiation and shielding, fireball growth and effects, thermal radiation, air blast, cratering and throwout, ground shock effects, fallout, and afterwinds. (J.R.D.)

The recovery of Th from Blind River ion exchange barrens with di(2- ethylhexyl)phosphoric acid was investigated. The recovery of Tc and Np from fluorination plant residues with tertiary amine was studied. The extraction of Np/sup 4+/ by quaternary ammonium nitrates is reported. A solvent recovery procedure involving successive use of Na/sub 2/CO/sub 3/washing and Al/sub 2/O/ sub 3/- adsorption was demonstrated in laboratory tests as a possible method for the purification and decontamaination of organophosphorus process solvents. The effect of nitrated fractions of Annsco 125-82 on Zr-Nb extractions by TBP was investigated. Treatnnent of TBPAmsco 125-82 solutions with 2 M HNO/sub 3/ at 60 ction prod- C for 1 to 48 hr showed that under these mild conditions the TBP degradation products were more important than those from Amsco as contributors to Zr-Nb extraction and as affecting efficiency of solvent clean-up. The interfacial tensions between benzene solutions of several amine salt and alkyl phosphate extractants and aqueous solutions were examined as functions of the solute concentrations. (For preceding period see CF-59-11-132.) (W.L.H.)

Recent borescoping of the KER-1 tube revealed several scratches, pits, and gall marks on the internal wall of the tube. These deformations could limit the operating temperature and pressure of KER-1. This report is a summary of operating history and is compiles to assist in determining what contributed to the condition of the tube.

The three-loop power system which is to be used with Experimental Breeder Reactor No. II (EBR-II) was analyzed to determine the coolant flow rate requirements at various power levels, coolant temperatures at various power levels, effects of heat exchanger sizes (system optimization), and effects of control errors. An intermediate heat exchanger, preheater, evaporator, and superheater are included in the EBR-II power system. Constant thermal resistances and physical properties, and perfect insulation were assumed in the analysis. Among other things, the study showed that at low power levels, excessive thermal stresses are produced at the cold end of the intermediate heat exchanger ualess a high sink temperature is used. It was also found that a short cut may be used to determine approximate system conditions at all power levels, that system optimization requires compromises, and that system flow rate control is possible through high-low measurements of two coolant temperatures in the primary system. (auth)

The scintillation counter has proven to be a very valuable research tool, but urfortunately, its ability to meet necessary stability requirements has restricted its use in industrial applications. Several techniques are being investigated which cancel out reasonable variations in detector sensitivity, resulting in improved stability. The general technique consists of alternately measuring the intensity transmitted through the sample and through a calibrated absorber, and difference in intensity causing the calibrated wedge to re-position itself. A comparison of commutating and noncommutating systems is made and other applications of scintillation counter systems are discussed. (For preceding period see ARF-1152-6.) (W.D.M.)

A capacitance instrument is described that indicates the level of liquefied gas in a closed container. The instrument has been used to indicate and control the level of liquid nitrogen, hydrogen, and methane. (auth)

The differential cross section for charge-exchange scattering of negative pions by hydrogen has been observed at 230, 260, 290, 317, and 371 Mev. The reaction was observed by detecting one gamma ray from the {pi}{sup 0} decay with a scintillation-counter telescope.

Based on a thesis submitted to Univ. of Kansas. Uranium monosulfide of high purity was prepared by reaction of uranium metal with hydrogen sulfide and subsequent homogenization of the crude product by heating at

An IBM-704 computer code for the harmonics method of criticality calculation for two-region spherical reactors is described. In the harmonics method, the criticality condition corresponds to the vanishing of a certain infiniteorder determinant; in practice, this condition is replaced by equating a finite-order approximating determinant to zero. By hand, the calculations can be performed conveniently only for second-order approximating determinants. The approximating determinant with the described code is customarily of the seventh order. Losses of significant figures prevented the use of larger determinants. The machine running time per case is generally about 30 sec. (auth)

This report discusses an outbreak of ruptures in January and February, 1960, which motivated serious consideration of ail phases of reactor operation so that a combined efforts to reduce rupture potential might be made. Items investigated include mixer location, gas composition control, rod movement, rod configuration, tube power, and the axial flux profile of the reactor.

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 capacity of a continuous solvent washer with critically safe dimensions was found to be 0.5 gpm when either 6% or 30% tributyl phosphate in kerosene was washed with a 1 M solution of sodium carbonate. (auth)

Two codes were written for the IBM-704 to calculate the behavior of a reactor following a step change in reactivity, using one-group. space- independent, zero-power kinetic theory. The reactor is assumed to be running at constant level before the step is made, either at critical or subcritical conditions, with an external source. The code RE-128 assumes all delayed-neutron precursors in equilibrium at the time of the step, while the code RE-135 allows cases with nonequilibrium precursors to be handled. RE-126 can handle the case of zero final reactivity. Both codes are written in FORTRAN language. (C.J.G.)

The High Flax lsotope Reactor (HFIR) is being planned for construction at Oak Ridge National Laboratory as a supporting facility in the program of investigation of the properties of the transplutonium elements. The reactor will be a flux-trap reactor consisting of a berylliumrefiected, light-water-cooled annular fuel region surroundin g a light-water island. An irradiation sample of 200 to 300 g of Pu/sup 242/ will be placed in the island where a thermalneutron flux of approximately 3 x 10/sup 15/ n/cm/sup 2//sec can be achieved on the average during an irradiation period of about 1 year. It is estimated that more than 100 mg of Cf/sup 252/ will be produced by such an irradiation. In addition to the central irradiation facility for heavy-element production, the HIKIR will have eight hydraulic rabbit tubes located in the beryllium refiector and four beam holes for basic research. Preliminary design of the reactor was based on the results of a parametric study of the dimensions of the island and fuel region, heat-removal rates, and fuel loading on the achievable thermal-neutron fluxes in the island and reflector. A research and development program ding critical experiments, heat transfer, corrosion, a clufuel element studies has been in progress …

A colorimeter is described that can be used to monitor process solutions continuously for uranyl nitrate or plutonium nitrate concentration. The instrument was tested under plant conditions in the concentration range from 0.1 to 70 grams of uranium per liter and 0.1 to 10 grams of plutonium per liter. The instrument error was plus or minus 1% of the span, but errors of 15 to 20% can be caused by other variables such as acidity and other salts present. (auth)