K-capture to positron branching ratios were measured in the decay of Na/ sup 22/, Co/sup 58/, and Ga/sup 68/ -- all pure Gamow-Teller emitters, using coincidence scintillation spectrometer techniques. The measured values are 0.105 plus or minus 0.004 for Na/sup 22/, 5.08 plus or minus 0.17 for Co/sup 58/, and 1.28 plus or minus 0.12 and 0.10 plus or minus 0.02 for Ga/sup 58/. From these the Fierz interference terra is computed to be b =--0.004 plus or minus 0.012, -0.004 plus or minus 0.14, --0-03 plus or minus 0.02, and +0.03 plus or minus 0.01, respectively. These results indicate that the Fierz interference in Gamow-Teller interaction is very small. The decay of 270 day Ge/sup 68/ was investigated in equilibrium with Ga/sup 68/ to look for a possible low-lying 0+ level in Ga/sup 68/ using x-ray--x-ray and x-ray- gamma -ray coincidences. The result was negative. Besides the 9 kev K x ray, the 1.07 Mev gamma ray in the decay of Ga/sup 68/ and annihilation radiation, no other gamma rays were detected (<8% of 1.07 Mev gamma ray). The number of positrons per 1.07 Mev gamma - quantum was determined as 19.47 plus or minus 2.10. The ratio of …

The equations of the Debye-Huckel theory, modified to include quantum statistics, are discussed. It is found that the nonlinear equations used by Cowan and Kirkwcod are not unique and that the nonlinear theory can be formulated in different ways to give different answers. The linearized equations of these alternative formulations are discussed, and the correct form of the linearized theory is established. From the linear theory, the Helmholtz free energy of a slightly degenerate plasma is derived, and from this result, useful formulas in the near-classical limit are obtained for the pressure and internal energy. (auth)

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Requirements for successfull polishing of metallographic specimens are discussed and the design of a machine to accomplish such work on radioactive specimens in a metallographic cave is presented. Quality results, readily obtrained by use of the ANL machine, are illustrated by photomicrographs. (auth)

A classical linear vector space is a unitary DELTA module, where DELTA is a division ring. Properties of linear spaces are given. The approach used is a module theoretic one, that is, a sequence of progressively stronger impositions on the module structure of a unitary Rmodule M is studied, without regard to the nature of the operator ring R. (W.D.M.)

The neutron flux distributions in an EGCR cell containing seven and clusters of 2.0 and 2.6a enriched uranium odde were obtained by using a one- velocity, one-dimensional P-3 solution to the neutron transport equation and adjusting fluxes in the fuel cluster in a manner which is consistent with previous comparisons of experiments and calculated distributions. Flux traverses in the outer rod perpendicular to diameter of the cluster are also presented. (auth)

Many elements exhibit anisotropic scattering at energies of interest in reactor calculations. A method is presented for the calculation of transport cross sections including the observed anisotropy. (auth)

Casting a U shield for a kilocurie Co/sup 60/ source is described. The melting equipment is described, and data on the castings are tabulated. Examination of the casting revealed no large blow holes or cracks, and the method was considered adequate. (J.R.D.)

Pu production from separation plants was only 65% of the monthly commitment owing to Purex difficulties. UO{sub 3} production and shipments both met schedules. Although unfabricated Pu metal production was reduced, all shipping commitments were met on schedule. Purex equipment responded satisfactorily to decontamination. 860,000 Ci of Ce{sup 144} were recovered from Purex Conc. IWW. The all-Ti L-3 concentrator loop was installed in the Redox Pu Concentrator. The safety of the slag and crucible dissolver in Finished Products Operation was improved by adding cadmium to each batch. Engineering studies of Palmolive facilities are reported. An emergency water supply for the Purex 241-A waste storage tank farm will be installed. A study was made on casks for NPR fuel shipment. (DLC)

The effect of nitrate on Th extraction by primary amines was studied. Selectivity of the primary amines can be greatly improved by adding alcohol to the kerosene diluent. The flow capacity of settlers for NaCl stripping in the Amex Process was determined in a 6-in. settler for alcohol-modified and unmodified 0.1 M Rohm and Haas LA-1 amine m kerosene A process scheme is proposed for the recovery of U and Pu from H/sub 2/SO/sub 4/ stainless steel decladding solutions, by extracting U/sup 4+/ and Pu/sup 3+/ or Pu/sup 4+/ with successive streams of 0.1 to 0.3 M primary amine and stripping ihe combined extracts with dilute HNO/sub 3/. The structural effects on Th and U extraction with neutral organophosphorus reagents were studied. A flowsheet was developed for the extraction between TBP and HNO/sub 3/. The kinetic behavior of amine extractants was studied. (W.L.H.)

The measured diffusivity of uranyl nitrate in water at 25 ction prod- C was 0.7 x 10/sup -6/ cm/sup 2//sec with about 40% average deviation. A program was started to develop nonnuclear uses for depleted uranium. Two continuous DRUHM reaction runs were terminated due to erratic operation of the sodium metering system. In the second Fluorox run with crude UF/sub 4/ which lasted for 29 hr, a total material balance of 94.8% was obtained and 17.9% of the theoretical amount of UF/sub 6/ was collected in cold traps and chemical traps. Room temperature flow rate-pressure drop calibrations of a multiclone (thirteen 0.60-in. diam hydroclones in parallel) for installation with the HRT replacement circulating pump were completed. Mixed oxides of U : Th = 0.08 : 1 and all have low yield stresses of 0.02 to 0.05 lb/sq ft compared to 0.2 to 1.0 lb/sq ft for normal Th-U or Th oxides of 1.5 to 2.5 micron mean diameter. The rates of uranium anion exchange from solutions containing between 0.025 and 0.20 M sulfate were measured and apparent uranium diffusion coefficients between 1.2 x 10/sup -7/ cm/sup 2//sec and 1.6 x 10/sup -7/ cm/sup 2//sec were calculated. In bench scale studies, …

Materials of construction were evaluated for use in critical areas of head-end processes for the aqueous reprocessing of spent nuclear fuel elements. The SulfexThorex, Darex-Thorex, Darex, Zirflex, and Zircex processes were considered. The effect of varying heat treatments on the resistance of the materials was also evaluated. Dissolution of unirradiated fuel pins was carried out in vessels of promising materials. The corrosion rate of Ni-o-nel was about 5 mils per month during actual fuel-pin dissolution by the Sulfex-Thorex process. Stabilization and heat treatment are necessary to prevent intergranular attack at welds. Carpenter 20 Cb is subject to stress-corrosion cracking by the Sulfex decladding solution and Illium R behaves similarly to Ni-o-nel in Thorex solutions. Titanium shows promise as a construction material for a Darex-Thorex dissolver. However, several questions remain concerning a vapor-phase attack observed around certain weldments. Carpenter 20 Cb, Ni-o-nel, and Types 309 and 309S Cb stainless steel appeared worthy of further study for the Zirflex dissolver. Preliminary evaluations show that at least Ni-o-nel and Carpenter 20 Cb should be studied further as possible construction materials for a single vessel for Zirflex and Sulfex-Thorex processes. Illium R, Hastellcy C, and nickel were not attacked by hydrcchlorination conditions of the …

Previous reports on the chlorination of plutonium dioxide describe numerous small-scale experiments and a few fair-sized batch preparations. The chemistry of chlorination by numerous reagents is covered, but no process had received sufficient study for large-scale preparation of anhydrous plutonium trichloride. The literature search revealed no extensive studies on chlorination rates, exhaust gas filtering, atmospheric requirements, reactor materials, etc. A program was undertaken to select a chlorination process, to develop the necessary information for defining operating conditions and equipment specifications, and then to demonstrate the operation of the process.

Results of an analysis to determine the extent of displacement of the EGCR core due to blowdown in case of several postulated hot main gas coolant pipe failures are summarized. Results show that the core will be damaged for ary hot pipe double-ended failure. Excepting the improbable case of no coolant flow existing prior to the break, the core will be damaged for any hot pipe fracture exposing a total flow area to the atmosphere equal to that of one pipe. Smaller breaks will probably be safe in this respect. (auth)

The rate of chloride removal varied directly with HNO/sub 3/ concentration fn an APPR-type Darex dissolver product containing 100 g/liter metal loading, 0.58 M initial chloride, and initial HNO/sub 3/ concentrations of 8, 9, 10, 12, and 14 M. The removal rate with 8 and 9 M HNO/sub 3/ was very low. After 6 hr refluxing, the chloride content decreased to 0.50 and 0.36 M, respectively. After refluxing for the same time with 10 to 14 M HNO/sub 3/, the product contained 0.064to 0.0007 M (2270 to 25 ppm) chloride. The effect of air sparging was approximately equivalent to refluxing without sparging at a HNO/sub 3/ concentration 2 M higher. After 6 hr sparging and refluxing the chloride content varied from 0.034 to < 0.00014 M (1200 to < 5 ppm) for initial HNO/sub 3/ concentrations from 8 to 14 M. (auth)

The decomposition of N/sub 2/O in a reactor tube containing various fixed-bed catalysts was investigated at 200 to 700 deg C, space velocities of 250, 1250, and 2500 vol. of gas per vol. of catalyst per hr, and various gas mixture compositions. As catalysts, En at 500 deg C and Pd at 650 deg C both gave satisfactory results. NO/sub 2/ was formed with all these catalysts, the amount increasing as the residual N/sub 2/O decreased. (C.J.G.)

Fifteen PWR blanket fuel specimens varying in burnup from 80 to 1100 Mwd/ T, were declad with boiling 6 M NH/sub 4/F-1.0 M NH/sub 4/NO/sub 3/ before the UO/ sub 2/ core was dissolved in 10 M HNO/sub 3/. Uranium and plutonium losses to the decladding solution were less than 0.2% in nearly all runs. While these ments with unirradiated fuel, they are of the same order of magnitude as those obtained in the testing of the hot cell HF was used as the decladding reagent, the uranium and plutonium losses averaged 1.0 and 0.4%, respectively. (auth)

The following criteria outlines the basis, objectives, and fundamental methods that shall govern the preparation of final design for ``Project CGI-839, Modification to Fuel Element Test Facilities -- 1706 KER.`` These modifications will provide the equipment to test NPR size fuel elements in the KER recirculating loops. The 1706-KER Recirculation Test Facility of KE Reactor is operated to obtain experimental data regarding high temperature reactor coolant technology and high temperature fuel element testing. The facility consists of four in-pile recirculating loops. These loops will permit testing of fuel elements with the existing process tubes of 2.1 inches I.D. To provide adequate in-reactor fuel element test facilities to support the development of NPR fuel, two KER loops, {number_sign}3 and {number_sign}4 will be converted to provide a process tube of 2.7 inches ID that will be operated at typical NPR irradiation conditions. The remaining loops No. 1 and 2, will be modified to provide additional flow and heat transfer capacity for greater flexibility in the testing of high temperature fuel elements smaller than the NPR size. New pumps, heat exchangers, and minor piping modifications will be required in all loops.

To evaluate the self-supported fuel element concept, tests are underway to determine the performance of collapsible bridge-rail supported fuel elements in ribless process tubes under present reactor conditions at B Reactor. It appears expedient, however, to extend this evaluation to future operating conditions in order to establish the relative feasibility of conversion to self-supported fuel elements in ribless tubes in present reactors. Utilization of 1.47% U-235 enrichment will provide fuel element powers comparable to those attainable under proposed future conditions. Since I & E fuel elements of this enrichment have previously attained exposures in excess of 2000 MWD/T at specific powers averaging 75 KW/ft in C Reactor, this test will specifically evaluate the feasibility of the self-supported fuel element concept. The purpose of this report is to present the design of a test to fabricate and evaluate self-supported fuel elements under conditions of comparable severity to those expected for future loadings of this geometry.

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Objective of this test is to authorize irradiation of alloyed, low hydrogen dingot uranium fuel elements on a pilot scale, and to monitor their performance. Initially, 25 tons per month of alloyed, low hydrogen dingot material will be charged for two months. Measured charges will be loaded with the initial 25 tons to monitor the stability of this material. Following a two-month delay in the monitor charging, and if the dingot meets all specifications, routine charging of quantities up to 60 tons/ month may proceed for six months and, assuming continued favorable performance, up to 150 tons/month may be accepted to complete large scale evaluation of dingot uranium, and on a continuing basis thereafter.

Heat loss tests were conducted with six insulation configurations for application in the riser regenerator and auxiliary regenerator sections of the loop. Insulation consisting of ten laminations of 0.003 in. stainless steel shim stock spaced 1/8 inch apart produced a temperature drop across the 15/16 inch annulus of 1200 F with a heat loss of 1.04 KW per foot of 2 inch schedule 40 pipe. The curve of heat loss vs. temperature difference is presented which, with results of similar tests with a 1/4 inch annulus, will permit the evaluation of a heat balance and temperature profile for the entire loop. (auth)

An air-cooled plugging meter for rapid determination of sodium oxide concentration in liquid sodium was tested in an experimental system. Approximately 200 plugging tests were performed, with results indicating good repeatability and a relatively fast operating time compared to other plugging meters. A typical operating time for making a determination with a system temperature of 725 deg F was 5 minutes. (auth)

Diffusion coefficients for hydrogen in beta zirconium were determined from permeation rates in the range 650 to 850 deg C. Both the steady-state method, which is dependent upon the hydrogen concentration, and the time-lag method, which is independent of hydrogen concentration, were employed to obtain diffusion data. Zirconium disks, 0.03 to 0.1 cm thick and varying in hydrogen concentration from 9 to 33 at.%, were used to measure permeation rates. The diffusion coefficients determined by the steady-state and time-lag methods on samples of differing thickness were in agreement. It was concluded that the permeation process was diffusion controlled. The diffusion coefficients were found to be independent of concentration and can be expressed by D = 6.14 x 10/ sup 4/ exp (--45,900/RT). (auth)