"The problem of radiation derange to the walls of two types of vacuum chamber for the 6 Bev Cambridge Electron Accelerator was studied. Radiation damage may arise from the synchrotron radiation and from electrons which are not accepted at injection. The synchrotron radiation produces a large amount of secondary x radiation which is isotropic, and therefore complicates the arrangement of internal shielding. The 20 Mev electrons from the lines cannot be attenuated by shielding. It is concluded that dosages of the order of 10/sup 7/ rad/day near the inflector are unavoidable. This would exceed the allowable dosage for a tube made by cementing a stainless steel liner to supporting ribs, and also exceed the test dosages (so far as we know) for alumina ceramics."

The purpose of the survey was to determine the radiation level in the Turbine Generator Services Building, around the Fuel Handling Canal, in the limited access areas of the reactor plant containers and on the boiler container roofs during plant operation. The test was performed with the plant at approximately 100 per cent power with three loops (1A, 1B, and 1C) in service. The radiation levels of the areas surveyed with a few exceptions were essentially background (.03 MR/HR). These exceptions were the Auxiliary Chamber and the limited access areas between the containers. The highest radiation level obtained in the survey was >2500 MR/HR at the reactor chamber walkway.

The mathematics has been developed for the determination of the radial coupling coefficient for a hollow cylinder of electro-strictive material whose length is small compared to its outside diameter.

This memorandum is a compilation of general information concerned with the arrangement of elements on a control panel. Arrangement considerations that lead to to improved ease and accuracy of operation are presented to assist the engineer in designing new control panels.

Fourier analyses are made on two functions. The first is a step function forward from periodic samples of a sinusoid. If the frequency of the sinusoid is less than one-half of the sampling frequency, it is shown that the step function has no frequency components less than one-half of the sampling frequency other than that of phase, and duration with respect to the interval of the analysis. It is shown that the insertion of a blank space in the period of analysis reduces greatly the uncertainty of the amplitude of the initial sinusoid as estimated from the results of the Fourier analysis. The results of the analyses are useful in the design and evaluation of certain analog data-analyzing systems.

A method for determining the concentration of Li6 in aqueous solution has been tested using the nuclear reactions Li6 (n,α)H3 and O16 (H3,n)F18. Annihilation 7 radiation of induced 1.87 hour F18 radioactivity was counted with a well-type scintillation counter, and the radioactivity per millimole of lithium was found to be independent of lithium concentration below about 0.2moles/liter. The sensitivity limit for detecting lithium is less than 0.1 micromole (0.0075 micromole Li6).

The purpose of this study was to develop equations for calculating fision product heating in the HRT-CP underflow pot from measured temperatures and to attempt to correlate the rat of solids accumulation in the underflow pot with fission heating and reactor power. Using fission heating data calculated from relating solids accumulation and heating have been tested. In one case an error of no greater than 26% was incurred in the calculation of the total weight of solids collected during chemical plant runs 17-4, 17-5, and 17-6. Further development work will be done on this correlation.

Various nuclear characteristics of two-region spherical homogeneous reactors have been computed in order to provide information for the design of HRE-3. Equilibrium isotope concentrations were established using an ORACLE code, and a two-group model was used to obtain critical concentrations and flux distributions. Breeding ratio is plotted as a function of reactor size, blanket thorium concentration, and other design and operating parameters, and the time required for a demonstration breeding is discussed. Tables of results, including neutron balances, are given for selected reactors. a number or relations are presented for estimating the effects of fission products, copper, corrosion products, H2O, and the core tank on breeding ratio.

Design calculations were made for a system consisting of a pump, one hundred feet of pipe, and a heat exchanger to remove 1 Mw of heat from various aqueous thorium oxide slurries. The rheological properties of the slurries were varied over a range of yield stresses from 0 to 1.5 lb/sq ft and of coefficients of rigidity from 1/2 to 2 centipoise. Two different cases were studied: a heat exchanger having fixed axial and radial delta T in which the tube length was allowed to vary and a heat exchanger having fixed tube length in which the axial and radial delta T were allowed to vary. It was shown that the pump power must be increased by a factor of 15 to 30 in order to maintain satisfactory operation of the heat exchanger as the slurry yield stress is increased form 0 to 1.5 lb/sq ft. However the pump power is essentially independent of heat exchanger tube diameter for any given slurry. The rated capacity of a slurry heat exchange is essentially independent of slurry yield stress and coefficient of rigidity, provided that the tube velocity can be suitably increased as the slurry yield stress in increased.

A falling ball viscometer under development at ORNL, employing a flow system and an electromagnetically operated dash-pot pump, was evaluated for possible use with aqueous ThO2 slurry systems under reactor irradiation. Interchangeable check valve inserts were designed and fabricated to investigate several pump designs. Magnetic flux concentrations which originally prevented circulation of the ferritic stainless steel viscometer ball through the electromagnetic pump were eliminated by substitution of iron-magnetic stainless steel inserts. Viscosity was correlated through a logarithmic plot of the dimensionless Reynolds number vs. Froude number with the ratio of ball diameter to tube diameter as a parameter. The relation is linear in the laminar flow region.

Some refined calculations have been made, relative to a proposed delay prior to a dump, to determine the expected D2 concentration in the vent stream from the pressurizer gas bleeds during a dump of the Homogeneous Reactor Test (HRT). These calculations indicate that for vent valves have a Cv of 0.07 (venting time from 2000 psia to D2O saturation pressure of approximately 12 minutes), a delay period is not required since the D2 concentration is well below lower explosive limit. For vent valves having a Cv of 0.3 (venting time approximately 2.4 minutes), the calculation indicate that a delay before venting of approximately two minutes will be required. This is due entirely to the possibility of mass ebullition the D2. Since the pressure drops so quickly, the reactor solution becomes saturated with D2 before appreciable recombination can occur.

A new type of continuous density meter, applicable for use with ThO2 slurry in high temperature-pressure systems, was tested successfully in a low temperature slurry loop.

Leaching tests were run on two samples from the Grants area in New Mexico. Uranium extractions of 94 per cent were obtained by leaching Sample 6-1 with solutions containing 240 lbs. of Na2CO3 and 60 lbs. of NaHCO3 per ton and by leaching Sample 6-2 with 270 lbs. of Na2CO3 and 180 lbs. of NaHCO3 per ton. Cyclic tests were completed using caustic precipitation of the leach liquor.

The ORNL Graphite Reactor operated very well during 1955. The downtime was low, only 8.6%. The fuel in the bonded slugs did not perform as well in 1955 as in 1954. Much of the trouble was undoubtedly due to growth of slugs which were not beta-transformed. It is known that some slugs had grown over 1/2 in. The automatic central system installed in 1954 continued to operate satisfactorily. The cooling system gave minor trouble when one of the 900-hp fan meters had to be replaced because of shorts in the rotor. The high radiation in the canal was the largest source of trouble. Approximately 55 tons of slugs discharged from the reactor in 1952 was sent to the Metal Recovery Plant. Enough slugs had raptured, due to their long exposure in the canal and reactor, to badly contaminate of water. Most of the contamination was removed by the end of the year, but the radioactivity which had soaked into the canal wells was enough to give high radiation fields. A solution to this problem was being sought at the end of the year. A study is under way on the possibility of increasing the flux of the ORNL Graphite Reactor …

Construction of the HRT reactor shield tank was completed, and the inside surfaces were painted. The roof structure for the tank is being assembled in preparation for an acceptance pressure test. Service piping and instrument lines are being installed in the central room area by ORNL craft forces. This work is approximately 50% complete. Fabrication of all temperature system components, except the blanket outer storage tanks, has been completed.

A preliminary chemical flowsheet is presented of a fluoride volatility process for recovering and decontaminating uranium from heterogeneous reactor fuels after dissolution in a fused salt. In laboratory work, a gross β decontamination factor of > 10 4 was obtained in the fluorination of a UF4-NaF-ZrF4 melt by passing the product UF6 through NaF at 650°C. The solubility of UF6 in molten NaF-ZrF4 was shown in kinetic studies to cause a lag in the evolution of UF6 from the fluorinator. Corrosion of nickel in the fluorination step appeared to be 2-4 mils/hr during the time that uranium was present. The average corrosion rate over the process as a whole was less than O.4 mil/hr. Earlier studies were reported in ORNL-1709 and 1877.

In a third run with the fluorine torch, the settling chamber wells were kept hotter than before (≥ 625°C); the flame was cooled by diluting the fluorine with helium. In the analysis of the products, 99% of the thorium fluoride fed in was accounted for, but only 64% of the protactinium activity. Part of this was carried in the exhaust gases past the cold trap and into the soda line disposal column, where it was detected by survey meters. The stripping of protactinium from the solid was somewhat more efficient than before; 77% of the feed which was recovered from the settling chamber had lost 72% of its original specific activity. About 15% of the input activity was trapped on the cold fingers with very little thorium fluoride.

The so-called Fluoride Volatility Processes refer to several proposed non-aqueous methods of processing irradiated fuel elements. In each of these methods, the uranium is fluorinated to UF6 and then decontaminated by distillation. One of those methods, involving the direct fluorination of the uranium by bromine trifluoride (BTF), has been under investigation at BNL since 1950. In 1952, it was demonstrated at BNL that uranium, as UF6, could be satisfactorily decontaminated by distillation in small-scale pilot plant equipment; end in 1953, BNL undertook the job of determining the technical feasibility of a continuous dissolver on a pilot-plant scale. The reason for the project was that the economic superiority of the process seemed to depend upon its amenability to continuous operation.

Considerable interest has developed in the use of ion-exchange in the nuclear energy field in the last decade. Aside from the obvious use of providing demineralized coolant water for reactors, the projected uses of ion-exchange include the recovery of fission products from aquaeous waste streams and the separation and purification of fissionable materials from spent reactor fuels. The latter process may be incidental to the over-all operation, as is the case with the Purex anion exchange facility, or it may be the prime separation process, as may be the case in the recovery of Pu or U from spent power reactor (PRTR) fuel elements.

Basic Studies. It has been reported previously that a reduction of PuO2 to a suboxide does not occur when a powder sample is heated for one hour at 1450 C. To investigate this anomaly, the present hooded facilities were converted from full air flow to an argon atmosphere to prevent oxidation of a possible suboxide. Five grams of PuO2 powder were heated in dry hydrogen to 1500 C for times of one and eight hours. Immediately after discharge, they were mounted and transferred to a helium atmosphere diffractometer hood. The resulting x-ray diffraction pattern consisted only of the single FCC PuO2 phase.

The basic process for fabricating a swaged fuel rod is simple, easy to control and inexpensive. A zircaloy tube is filled with uranium dioxide powder, the ends temporarily plugged and the loaded tube is swaged to compact the UO2 powder to the required density. The swaged rod is then cut to length and counterbored and then end cape are welded into each end. After several tests and inspections, nineteen rods which meet the quality standards are assembled into a single fuel element ready for irradiation.

The Zircaloy-a clad, swaged UOa, 19-rod cluster fuel element for the PRTR was designed to use Zircaloy-a wire spirally wrapped around the fuel rods as spacing members. Such use of unbonded, Zircaloy-a spacers introduced the possibility of fretting corrosion. This paper reports preliminary irradiation tests conducted to determine whether or not such corrosions occurs in this fuel element design.

In an earlier study on criticality conditions for homogenous mixtures, 2/cm^2 was used as the neutron age for all mixtures of water and uranium. At the higher H/U ratios (low uranium concentration), the calculated critical parameters were in good agreement were in good agreement with experimental data. At the low H/U ratios (high uranium concentrations) the calculated critical parameters were smaller than the experimental ones (more conservative from a nuclear safety point view). These results indicated that using 27 cm^2 as the neutron age gives increasingly conservative results as the H/U ratio decreases.

The measurement of a alpha-particle energy has been used by many radiochemical laboratories for the identification and analysis of alpha-active radio nuclides. The use of the total-ionization method for alpha-active radio-nuclides. The use of the total-ionization method for alpha energy in ionization chamber in which the alpha particle loses all its energy in ionization of the chamber gas. Collection of the electrons thus formed generates a voltage pulse across the chamber capacity which is proportional to the alpha particle energy. This pulse is then amplified using a suitable linear amplifier and fed to a pulses as to amplitude; the information is then recorded or stored. Since the pulse amplitude is proportional to the alpha energy lost to the chamber gas, the pulse height analysis can be used to estimate the energy of the alpha particles and in the case of several alpha emitters of different energies, the relative abundance of the alpha emitters can be determined. An alpha energy analyzer system using the ion collection method has been fabricated for use in radiochemical laboratories required to perform a large number of alpha energy determinations. This report describes the operation, maintenance, and application of this alpha energy analyzer system.