The neutron tissue dose at large distances from a fission source was studied by using a water-filled phantom and four different detectors: a BF/sub 3/ counter, a polyethylenelined ethylene-filled proportional counter, indium foils, and nuclear emulsions. The source of fission neutrons was the ORNL Health Physics Research Reactor which was attached to a hoist which was in turn installed on a 1530-foot tower. The reactor could be operated at any elevation from 27 to 1500 ft. The phantom studies were made at horizontal distances from 250 to 1500 yards from the tower. Dose contributions from recoil protons H/sup 1/ (n, gamma )D/sup 2/ and N/sup 14/(n,p)C/sup 1 reactions are considered. (auth)

In two previous reports nucleon-meson cascade calculations were carried out for several cases of interest in the design of the transverse shield for the proposed 45-Bev linear electron accelerator at Stanford University. Results are now given for two additional cases. (auth)

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This document summarizes the amounts of radioactive contamination discharged to ground from separations facilities through December 1962. Detailed data for individual disposal sites are presented on a month-to-month basis for the period of January through December 1962. Previous publications of this series are listed in the bibliography and may be referred to for specific information on measurements and radioactivity totals prior to December 1962. Tables list the major disposal sites in the separation facilities, total volume of waste discharged to each location, and the gross amounts of plutonium and beta particle emitters discharged to ground since startup. This same data is presented on a monthly basis for cribs still in use. Information is presented on the source of the waste stream and the settling facility if used. Isotopic data are included for disposal sites from which the waste was analyzed for specific contaminants. Estimates of contamination and volumes discharged to swamps are also included.

Volatile compounds containing C{sup 14} produced in the cooling gas systems of Hanford reactors can be lost from the cooling gas by two routes. These are (a) escape via the reactor stacks through leaks in the cooling gas system and (b), adsorption with water vapor on the silica gel of the dryers for the cooling gas with subsequent regeneration of the silica gel and condensation of the water for disposal. Information relative to the release rates by the stack gas route was reported in HW-75618. The present report provides data relative to the C{sup 14} loss rates through the condensate obtained from regeneration of the silica gel dryers of the reactor gas cooling systems. The C{sup 14} containing compounds in the reactor gas are produced by n,{gamma} reaction on the nitrogen in the coolant gas, an n,{gamma} reaction on carbon in the CO{sub 2} component of the coolant gas, or an n,{gamma} reaction on the graphite moderator with subsequent volatilization. Studies are presently under way to determine the contribution of graphite ``burnout`` to the C{sup 14} containing volatile compounds in the coolant gas. Upon completion of these studies a topical formal report will be Issued which will include the findings …

The possibility of using thermal-neutron radiography for inspecting heavy metals and hydrogenous materials was examined. The data include exposure curves, contrast sensitivities, and an assessment of the influence of higher energy neutrons and interfering gamma radiation on image quality. It is shown that, in the case of homogeneous materials, neutron radiography presents definite advantages for the inspection of heavier metals such as uranium, bismuth, and lead, and that the images obtained in such inspections are influenced very little by other radiation in the thermal-neutron beam. This is somewhat less true for intermediate metals such as steel and tungsten, although in these cases too, some exposure time advantage can usually be gained. Nevertheless, neutron radiographic inspection of these intermediate materials may be limited to those cases in which some complication, such as radioactivity of the sample, is involved. Thermal-neutron inspection of hydrogenous materials having a thickness greater than about an inch is not recommended and may be useful primarily in special cases such as one in which the light material is in some combination with a heavier, x-ray absorbing material. (auth)

Tributyl phosphate can be used for extraction in processing all current power reactor fuels. Nitric acid is the only salting agent required. Typical flowsheets are presented. In aluminum nitrate systems which are more than 0.1 M acid deficient, the uranium distribution coefficient is a function of pH and independent of aluminum concentration; the coefficient remains constant at one in fluoride systems when the nitrate to fluoride ratio is approximates 3.5. Many objectionable properties of degraded diluents are ascribed to nitroparaffins. Aliphatic diluents with the least branching are the most stable to nitration. The nitration stability of aromatic diluents varies with structure, e.g., stabilities of diethylbenzenes decrease as meta >> ortho > para. Solvent purification by flash distillation appears superior to other methods. The stability of Amsco 125-82 was permanently improved by treatment with sulfuric acid. The radiation stability of TBP was approximates 2 times higher in an aromatic diluent than in Amsco 125-82. The G decomposition value for 1 M TBP in Amsco alone was approximates 0.9; whereas in 1 to 3 M HNO/sub 3/ it was 1 to 5 and G (--HNO/sub 3/ org phase) was 3 to 20. Variation of uranium--thorium separation factors with structure of some neutral …

Phase acceptance of the Alternating-Gradient Synchrotron Linear Accelerator is theoretically examined and found to agree well with an experimental determination of phase acceptance. Bunching in the AGS linac is also analyzed, the results indicating a satisfactory understanding of this process also. A double buncher that should be more efficient than the present single buncher is discussed, and space charge effects (as yet unobserved in bunching) are analyzed. (D.C.W.)

The WPE progress report for September, 1962 documents reasons for suspecting nitrogen as a possible cause of weapons fabrication difficulties and loss of production caused by radio-graphic rejects. This is a summary and documentation of the ``Nitride Problem`` and related subjects concerning fabrication metal. It contains the background information leading to details published in the Weapons Process Engineering Progress Reports. This summary may provide a better understanding of the problem, since little of the background was suitable for publication in the WPE progress reports.

The problem of coupled radiation and convection to a medium in slug flow between infinite, parallel plates was formulated in terms of discrete fluxes. Steady state, uniform heat input, and gray boundaries were assumed. The effects of absorption, emission, scattering, and flow were taken into account. For the general case where all these effects are present, it was necessary to approximate the fourth order temperature term by a truncated Taylor series. This approximation was not necessary for the special case where conductivity of the medium is negligible. Approximate, analytical solutions in closed-form for profiles of temperature, radiant fluxes, and heat fluxes were obtained for both the special case and the general case. Sample results are presented, showing the effects of convection, absolute temperature, absorption coefficient, scattering coefficient, and wall emissivities. The two cases of equal wall temperatures and unequal wall temperatures were both examined. It was found that for a fluid with the thermal conductivity of air, radiation can account for 60 to 95% of total heat transfer in the temperature range of 900 to 2800 deg R. It was also found that for some situations, the scattering phenomenon can have a much greater effect on heat transfer than the …

Batch denitration of synthetic Purex waste was investigated in 12 batch runs in the pilot plant denitration unit. Sugar was continuously added to 25 liters of hot waste. After sugar addition was complete, the hot solution was digested for several hours. The reaction proceeded smoothly and was easily controlled. About 19 to 22 moles of nitric acid were destroyed per mole of sugar with a reaction pot temperature of 100 deg C and a digestion period of 12 hours. About 17 moles of nitric acid were destroyed per mole of sugar when the reaction pot temperature was reduced to 95 deg C. A possible plant flowsheet for batch denitration with sugar was developed. This ilowsheet presumes batch denitration of large 2,500-gallon batches of waste in a standard 5,000-gallon Purex Plant tank and incorporates 12 hours of sugar addition and 12 hours of digestion. The flowsheet was successfully demonstrated in the pilot plant equipment. A 1.4M sugar solution was added to 25 liters of hot (100 deg C) waste for 12 hours and the mixture was digested for 18 hours. The initial nitric acid concentration was 6.14M. The residual nitric acid concentration was 0.94 and 0.90M after 12 and 18 hours …

The isotope Re/sup 189/ was produced by fsst neutron irradiation of osmium and by the ( alpha ,p) reaction on tungsten. The rhenium was separated chemically from the target material, and beta, gamma, and internal conversion spectra were measured. The mass assignment is confirmed by the observation of eleven electromagnetic transitions in the Os/sup 189/ daughter, including the 30.8-kev isomeric transition (8 hr), all of which were known from the decay of Ir/ sup 189/. Re/sup 189/ has a half-life of 23.4 plus or minus 1.0 hr and emits betaray groups with end-point energies 1000, 780, and 725 kev, and probably others. Results of coincidence measurements lead to some new information about the level scheme of Os/sup 189/ Evidence was found for strong rotation-particle coupling between low-lying K =1/2 and K = 3/2 bands in Os/sup 189/. 40 references. (auth)

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Using six hot-wire columns and three concentric tube thermal diffusion columns a cascade system was arranged to produce 0.04 g of 90% C/sup 13/ per day using methane as the feed gas-. The ten liter shipment of purchased methane enriched to 60% C/sup 13/ was received. This gas will be enriched to 90% C/slup 13/. Before removing the equilibrated gas from the concentric tube columns in preparation for using methane enriched in C/sup 13/ as feed material, a sample of methane taken from the bottom of the last column contained greater than 65% C/sup 13/. The hydrolysis of protactinium was found to be completely controllable and reversible and can serve as the basis for its separation from other elements, particularly its decay products. The tendency to hydrolyze is not a unlque characteristic of protactinium, but is exhibited, under appropriate conditions, by Th/sup 227/ and Ra/sup 223/ at trace levels. Traces of protactinium in 1N H/ sub 2/SO/sub 4/ can be removed from solution by a coarse glass wool filter through which the decay products pass readily. The reverse is true if the H/sub 2/SO/sub 4/ concentration is increased to 6N. The desorption of protactinium from platinum was a function of …

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A reinforced concrete-shielded enclosure intended to contain nuclear reactor experiments was subjected to internal pressures up to 10 psig. The strains in reinforcing steel were measured with strain gages at critical locations for each pressure increment. These strains were compared with calculated strains; good agreement was found if the appropriate stress- concentration factors are applied at locations near re-entrant corners and wall openings. Leakagerate tests were performed to disclose the ability of an ordinary thick-walled concrete cell to contain compressed gases. Observed leak rates are given for the two twin cells, before and after the inside surfaces were painted. (auth)

To evaluate the crudding characteristics of ammonium hydroxide conditioned coolant it is necessary to use a thermocouple element. The present design has accentuated structural integrity, reliability and constant heat flux. A radical departure from past designs was made by selecting a stainless steel cladding and enriched ceramic fuel. Three thermocouples are inserted in the cladding to increase the reliability. By the use of two thermocouple elements it will be possible to obtain important preliminary information on preferential crud deposition while simultaneously evaluating the crudding characteristics of ammonium hydroxide conditioned coolant.

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Desirable characteristics of pulsed magnets used as ejectors or inflectors for large accelerators can be obtained by loading the magnet with distributed capacitance or by using a capacitor in shunt with the load. Comparable field build-up tlmes are available with either arrangement. The lumped magnet system offers greater simplicity. (D.C.W.)

The length of the minimum outage, or period of time during which a reactor must remain subcritical following a shutdown from equilibriu, is a direct function of the available excess reactivity. In reactors with unlimited safety system capacities, it is quite feasible to ``build in`` sufficient excess to allow complete override of the shutdown xenon transient, and thus, in effect, reduce the ``minimum downtime`` to zero. Present Hanford production reactors do not possess this unlimited capacity, and thus total pile reactivity must be restricted by minimum safety strength conditions. Any effort to increase the pile reactivity must therefore be tailored to satisfy existing Total control and Speed-of-control criteria. Production Test IP-467-C. ``Reducing Minimum Downtime,`` was designed to evaluate the economic benefits of a system in which base reactivity can be increased in periods of minimum control requirements, and correspondingly decreased when reactivity or safety considerations become more demanding. The purpose of this document is to summarize initial operating history and results obtained under the production test.

A proposal has been made to achieve a greater overall total production from the N-reactor through multi-product operation. Apparently, this can be achieved by producing tritium along with plutonium. Tritium production technology is presently based on low temperature and pressure water environments at SRP using targets which contain Al-Li alloy. Past experience has shown that the transport of tritium and lithium and the swelling which can result from helium agglomeration are all strong functions of temperature. Since operation of target elements at N-reactor conditions constitutes an extrapolation from available operating experience, a R&D effort is required in order to delineate and resolve any problems associated with target element operation in the N-reactor. This proposed irradiation test is a significant part of this R&D effort.