The pneumatic temperature probe (PIM) is a device for measuring gas temperatures by utilizing the dependency of the flow of gases through a restriction on the temperature and pressure conditions. The determination is made by measuring critical mass flow across a restrictive element such as a nozzle and by knowing the upstream pressure and other variables pertinent to the critical flow equation, computing the temperature at the entrance to the restrictive element. In practice it has been found useful to use two critical flow nozzles in series and measure the ratio of the pressures at the nozzle inlets, together with the temperature at the downstream nozzle. The limitations of present thermocouple materials for long term use at elevated temperatures makes resort to this device attractive on many installations.

The steady-state radial and axial temperature profiles of an E. G. C. R. 5 1/2" O. D. through tube are determined for the test condition of an attemperated fuel assembly operating at 1500 KW in the loop. The profiles are determined for the case of the central control rod fully inserted and bank insertion to 62 inches (Δk = 0.025). The data are presented in the form of tables and curves.

After shutdown heating rates have been calculated for the target, control plates, and the beryllium reflector of the HFIR. Hilvety previously calculated the after shutdown heating rates in the element and his results are reported in ORNL CP-60-4-110. Decay times of 1.0 10, 10^2, 10^3, 10^4, and 10^5 seconds have been considered, and heat fluxes have been computed for all of the mentioned components. The greatest heat fluxes were found to be at the surfaces of the control plates and the permanent beryllium reflector.

A chemical flowsheet is proposed for recovery of strontium and rare earths from Purex wastes. The iron in the waste is first complexed with tartrate and the pH is adjusted with caustic. Strontium and rare earths are extracted by a solvent comprised of di(2-ethylhexyl)phosphoric--sodium di (2-ethylhexyl) phosphate--tributyl phosphate -- Amsco 125-82. and are stripped with nitric acid. Processing the strip product through additional solvent extraction cycles yields isolated, concentrated strontium and rare earth fractions.

Five irradiated and five unirradiated wafers were analyzed. Each wafer was analyzed for samarium by emission spectrography. The unirradiated wafers were analyzed for uranium by coulometric and potentiometric methods and for uranium isotopes by mass spectrometry. The irradiated wafers were analyzed for uranium and plutonium by coulometric methods, for plutonium isotopes by the 256-channel alpha pulse analyzer, and for the isotopes of uranium and of plutonium by mass spectrometry. The methods of preparing wafers for analysis are discussed; the data are tabulated.

The feasibility of improving the mechanical properties at 1700-1800°F of oxidation-resistant Fe-Al-Cr alloys by means of a refractory dispersion has been explored. A literature search was conducted, preliminary experimental determinations of properties of the alloy and its oxides were carried out, and certain mathematical relations between dispersion characteristics and metallurgical variables were derived. The results indicate that the alloys can be strengthened sufficiently by using a dispersion with an interparticle spacing of about 2-3 µ. High-temperature native oxides of the Fe-Al-Cr alloy consist largely of Al2O3 and in theory would serve as a satisfactory second phase.

Report that describes the Oak Ridge National Laboratory Experimental Gas-Cooled Reactor, problems with the procurement and assembly of its components, and its economic feasibility.

Report containing ongoing projects and experiments undertaken by the Oak Ridge National Laboratory's Molten-Salt Reactor Program.

Program using the IBM-704 computer to calculate certain kinematic quantities for any non-relativistic reaction of the form.

The thermal-neutron flux prevailing near the upper hole in the HRE-2 core wall was determined from measurements of the induced activity in a specimen cut from the edge of the hole. The specimen was removed following run 21 and was thus exposed to reactor neutrons during runs 13, 14, 16,17, 18. 19, 20, and 21. The neutron flux, calculated for an average core power level of 2 Mw, was 2.8 x10^13 neutrons/cm^2 sec. the core wall area from which the specimen was removed was located approximately 16-in. from the core center.

The ORNL analog computer was used to examine one aspect of a safety system for the EGCR. The thermal behavior of the fuel as a result of increases in reactivity at various power levels and gas flows was examined. Safety actions initiated by wither a flux-minus-flow signal or a reactor outlet gas temperature signal were included in simulation.

Results of processing 47 metric tons of Brookhaven spent reactor fuel are summarized.

Multigroup one-dimensional calculations were done recently to obtain estimates of critical masses, power density distributions and fissioning spectra for some homogeneous molten salt reactors having outer reflectors and central "islands," placed inside the currently proposed MSRE vessel. For a 5-inch-thick outer reflector and 1-ft-diamter island, both beryllium, the calculated critical mass is 108 kg; 40 percent of the fissions occur at thermal, and the maximum power density of 3.9 times the core mean power density occurs at the island-salt interface. If the reflector thickness is increased to 10 inches, the critical mass is reduced to 34 kg; 67 percent of the fissions occur at thermal, and the peak power density of twice the core mean again occurs at the core island-salt interface.

In the Communications of the ACM of April, 1960, p. 240, there appeared two Algol algorithms for the computation of In, the modified Bessel function of the first kind. One of these algorithms uses a convergent series and the other an asymptotic series. Their author, Dorothea S. Clarke, did not give any information with regard to which algorithm is more appropriate in a given situation (a given situation meaning the computation of In (x) for a definite pair of numbers n and x). Such information is necessary, however, if the algorithms are to be useful, particularly since the asymptotic series is in many cases not valid.

The hardness of several materials that have been considered for use as valve scats and bearings in thorium oxide slurry systems were measured and are reported for comparison with thorium oxide.

These specifications for the fuel assemblies for Core I of the Experimental Gas-Cooled Reactor (EGCR) have been developed for use in procuring the first core loading. A fuel assembly for EGCR consists of a cluster of seven cylindrical fuel elements spaced and supported within a graphite sleeve. Each fuel element consists of a stainless steel tube containing a column of hollow UO2 pellets and having a spacer brazed at the midsection to control the spacing between fuel elements in a cluster. A master specification for the fuel assembly, a supplementary specification for each of the components, and a specification on record keeping during manufacture are included.

A helium coolant purification system was designed for the proposed 800 MWT Pebble Bed Reactor. The purification system will operate on a coolant side stream with a flow rate 1% of the total coolant flow and there are provisions for radioactive and non-radioactive contamination removal.

Four naturally occurring materials, namely, vermiculite, variscite, Tennessee rock phosphate, and Florida pebble phosphate, were studied to determine the feasibility of their use in columns for the sorption of strontium from high pH, intermediate-level wastes produced at ORNL. Excepting vermiculite, all materials were found to be effective in strontium removal.

An estimate was made of the beam hole currents to be expected from both the radial and tangential beam holes in the present HFIR design. The angular fluxes required for these calculation were determined by an 18-group, S8 transport calculation (SNG reactor code). Both type beam holes produced thermal neutron currents of 7.5 x 10^10 neutrons/cm^2 sec. at a power level of 100 Mw. The nonthermal component of the beam current was 7.0 x 10^10 and 4.5 x 10^10 neutrons/cm^2 -sec for the radial and tangential holes, respectively.

A generalization of Wigner's simple model for the distribution of nuclear level spacings is studied. It is based on a stochastic process which reproduces the correct joing probability distribution of N enerfy levels for small spacings. The case N=3, which includes the effect of the correlation between adjacent spacings, is discussed in detail. The resulting distribution and the correlation coefficient are compared with experimental data. No definite conclusion can be drawn except to say that the effect of the correlations on the spacing distributions on the spacing distribution is very small.

The development of nuclear reactors and other nuclear particle sources has given the analyst a new analysis method which can be successfully applied to the determination of microgram and submicrogram quantities of many elements. Known as "radioactivation analysis", this method is one in which an "activation" by some type of nuclear reaction is used to produce a radioactive isotope of the element to be determined. Since this radioisotope decays with its own characteristic radiations and half-life, it is possible to make radioactivation analysis a very specific analysis. Chemical separations of the radioisotope are employed whenever necessary and its radioactivity measured by some type of radiation counter.

An analog simulation is made of the HRP in-pile slurry loop in the LITR. The radiolytic gas pressure in the pressurizer is determined. This pressure is a result of the generation of radiolytic gas in the loop core. A graphical result is obtained showing the effect of catalyst activity and pressurizer flow rate on the pressure rise from radiolytic gas. The thermal behavior of the system is studied, and the response to various controller settings is predicted. Controller settings very near optimum for the actual process were recommended from the analog study,

The hazard of Sr^90 arises in part from its fixation by phosphates in the skeletal system. It is natural then to consider compounds and minerals associated with the phosphate anion for applications in waste disposal. Calcium phosphate floc formation is widely used as a method for removing strontium from contaminated water. Also, calcerous soils are said to be effective for removing strontium from low-level wasters containing dissolved phosphates. It is suggested that one the mechanisms involved in strontium removal is the formation of calcium phosphate (apatite), with strontium substituting isomorphously.

Equipoise-2, a two-dimensional, two-group neutron-diffusion code in R-Z geometry, has been programmed for the IBM-7090 computer. This code was designed to permit the running of large numbers of cases without requiring excessive machine time. Typical running times are of the order of one to three minutes per case for a 1000-point problems. The maximum number of mesh points that can be used is 1444.