Pressure drop calculations are shown for He cooled power plants ranging from 400,000 kw to 30,000 kw.

To determine whether or not the (n,2n) reaction in Be would produce a net increase in fission neutrons in a power plant, the distribution of In resonance neutrons slowed down from fission energies in a graphite block was measured with and without a 2" of Be in front of a U3O8 slab undergoing fission. The thermal neutrons producing the fissions were obtained by slowing down neutrons from a cyclotron source. The observed distribution without Be was well represented by a Gaussian source range 36 cm. (corresponding to an initial fission energy of 3 MeV) and a similar sink of range 7.8 cm. The total In resonance intensity with Be was 5% less than without Be. On the assumption that the effect of the Be is entirely due to its different mean free path and moderating power, the In distribution with Be was calculated. Since the m.f.p. as a function of energy is not known for Be, two separate calculations were made using the highest (2.9 cm.) and the lowest (1.80 cm.) possible values for the effective Be m.f.p. Both calculated curves were found to be higher than the observed Be distribution, indicating that at least 10% of the fission neutrons …

The critical mass of 94-239 and the corresponding critical dimensions of homogeneous mixtures of 94-239 with various moderating media been calculated as a function of the concentration of 94. A simple transformation makes the figures applicated to92-235. the results are in essential agreement with the preliminary estimated made independently by Oppenheimer and Serber. The problem of the stability of a chain reaction in solution and questions of protection are discussed.

The yield of 49, the efficiency of production of 49, and poisoning in a power plant are discussed. Only the crudest of estimates of the poisoning are possible: these indicated that production will probably not be hampered by poisoning. In this case the yield of 49 could be as high as 3 kg/ton but only about 2 kg/ton is compatible with a fairly high efficiency. In the case that production is stopped by poisoning, smaller yields, proportional to the tolerable loss in k, are obtained. In this case the yield will be improved by a factor of 2 or 3 if only the most poisoned parts are extracted and replaced by new uranium.

The multiplication factor and minimum pile size for a multiplying pile using UO2 powder of density 1 have been computed. It appears that a k of 1.0177 is possible for a volume ration of V/V001=3.33.

The average energies of neutrons emitted from a graphite column at 22 degrees C were compared by measurement of the cross section of boron for neutrons which are stopped by cadmium. At a distance from the neutron source great enough to insure that the neutrons were in thermal equilibrium the average energies of the emerging neutrons were found to be proportional to the temperature within the limits of the experimental error. A measurement made with boron absorbers which had been thus standardized in the graphite column indicated neutrons emerging from the chain reacting pile to have an average temperature approximate 60 +- 50 degrees above that of thermal neutrons emerging from the graphite column at 22 degrees C. Such a measurement made inside the chain reacting pile indicated the average temperature of neutrons therein to be about 65 degrees +- 15 degrees above the average temperature of neutrons in the graphite column.

The most important source of radioactivity at the exit manifold of the pile will be due to O19, formed by neutron absorption of O18. A recent measurement of Fermi and Weil permits to estimate that it will be safe to stay about 80 minutes daily close to the exit manifolds without any shield. Estimates are given for the radioactivities from other sources -- both in the neighborhood and farther away from the pile.

The temperature coefficient is calculated for various lattice arrangements, taking into account the variation of [formula], suggested by Fermi. Four contributions are included: leakage, levelling of the dip in thermal neutron density in the lump, resonance absorption, and hardening of the neutrons as they penetrate a metal lump. The departure of neutron temperature from lattice temperature decreases the total coefficient. Values are given for 3 typical piles; in general, the larger the uranium elements, the less stable the pile. A rod lattice tends to be more stable. A pile with metal lumps over 50 lbs. will be unstable.

The resonance absorption of uranium for neutrons has been investigated between 20 degree C and 1000 degree C. Experiments were caried out on both UO2, density 4.63, and metal. The resonance activity was measured with respect to that of an iodine monitor at several different temperatures and the ratio of activity at temperature T to that at 20 degree C was determined. The increase in activity is 0.9 per cent per 100 degree C for the oxide and 1.1 percent per 100 degree C for the metal. The period of U239 was found to be 23.54 +- 0.05 min.

The importance of fast neutron fission (i.e., fission caused by neutrons before being slowed down) was recognized by Szilard, and calculations similar to the present one have already been carried out by him, Feld, Ashkin, Wheeler, Wigner and others. The purpose of the present paper is to give a general formula for the contribution of fast fission to the multiplication constant, which will include all the cases already considered and will be applicable to more complicated geometries than those considered by the previous writers.

The following represents a summary of calculations on the multiplication constant of homogeneous mixtures of uranium and different moderators. These calculations were made possible by Fermi's determination of the age of neutrons and by the extrapolation to higher scattering cross-sections of the resonance absorption of uranium as measured by C. Creutz. According to Fermi, the former quantity is 120 sq. cm. The latter is given in the two attached graphs. The first (Fig.1) of these goes as high as a scattering cross-section of 70 x 10-24 cm.2 per uranium atom, and its highest point is taken from a measurement of Cruetz's in which a mixture of U3O8 and graphite was used.

The following sources of contamination in cooling water of the P-9 plant are considered: (1) Fission recoils, (2) Corrosion of metal, (3) Recoil from aluminum, (4) Induced activity in the water. It is found that for a P-9 plant of 3.5 x 10(4) KW contamination of the river at "X" should not exceed the .1 r criterion unless coating failure occurs. Tables of the amount of coating failure permissible are given as a function of holdup time.

A model for the capture spectrum of uranium is introduced in which levels occur at 7, 30, 30+D, 30+2D, ..., ev. Gamma ray and neutron widths are derived from the experimental data for values of D between 10 and 25 ev. The constants of the 7 volt level prove to be independent of D. Extrapolation, using the one level formula, gives a capture cross section at (1/40) ev of 4.9 x 10(24) cm2; this number is, however, quite sensitive to the value taken for the resonance activation. Both neutron and gamma ray widths for the higher levels are approximately proportional to D.

Radial flattening of activity in the cores of spherical and cylindrical piles is discussed in connection with pile control and power improvement. Partial flattening as a result of k loss from temperature rise is also considered.

Some idealized considerations of the temperature field in the homogeneous pile are given. It is crudely estimated that the effective mean temperature rise of the pile to be used in calculating the k loss is something like 3/4 the rise in temperature of the slurry in passing through the pile.

Six structural materials (aluminum 2S, aluminum 72S, copper, iron, steel, lead and bismuth) were irradiated for about 100 kilowatt hours in the center of the Argonne pile. The alpha and gamma decays were then followed for 130 days on Geiger counters and are presented here graphically as number of rays per minute per gram per kilowatt hour vs. time of decay. A quantitative interpretation of the gamma-ray decay curves gave fair agreement with the chemical analysis of impurities found in the materials.

The multiplication constant and optimal concentration of a slurry pile is recalculated on the basis of Mitchell's recent experiments on resonance absorption. The smallest chain reacting unit contains 45 to 55 m3 of D2O.

The temperature coefficient of eta has been measured with the thermal column of the Argonne pile using uranium foils of different isotopic ratios. The temperature change was effected by filtering neutrons through silver. The measured fractional change per degree centigrade is [formula].

Some methods for calculating the k requirements of various shaped piles are indicated, with particular reference to the shape of the tank in the homogeneous plant.

The present report is concerned with the electronic relay which, upon the reaching of a pre-determined value of ionization current in the associated chamber, actuates the safety rod release mechanism. The are here proposed two independent suggestions: (1) An improved trip circuit of such sensitivity and stability as to allow it to be actuated by a signal directly from the chamber without the necessity of interposing intermediate amplifier stages. The value of chamber current at which tripping occurs is stable to +- 0.4%. (2) A triple-wound relay to prevent shut-downs of the pile which would otherwise be caused by development of defects in the safety circuits.

Calculations were made to determine the temperature at various points of a thick disc and cap, on the assumptions that the heat production rate is uniform throughout the slug and that the slug is in perfect thermal contact with the cap. Results obtained by different methods of computation are compared with each other and with results obtained by Drew on the Paschkis electrical network computer at Columbia University.

The resonance absorption of a lump is described in terms of three components. Calculations of resonance absorption are carried out using the model for the resonance spectrum of uranium recently derived; calculations are also made with variations of this model which involve local fluctuations in level strengths. For metal and oxide lumps the agreement with observation is satisfactory. For dilute mixtures, whose resonance activation was measured by Mitchell, computed values fall 20% to 30% above measured ones.

An attempt is made to estimate the influence of operation upon the mechanical properties of tubealloy. It is to be emphasized that the work is highly speculative, being base upon a set of judicious assumptions regarding the transport material within the metal and should not be used as a substitute for a set of well-planned experiments. The two principal effects considered are the hardening which results from the atomic displacement produced by the fission fragments and the changes in mechanical properties arising from the presence of the fission products. The principal conclusion to be drawn is that the disruption of the metal which accompanies operation takes place at at sufficiently high rate that it could readily cause serious changes in the mechanical properties of the slups. the reversal of the disruption resulting from thermal effects probably is sufficiently high to prevent serious embrittlement as a consequence of displacement through most of the volume of the slug if not at the surface. On the other hand, the migration of atoms also appears to be sufficiently high that the rare gas products may have time to diffuse to cracks and produce embrittlement.

It is usual to determine the density of neutrons by using foils of metals which become radioactive as a result of bombardment in the region under investigation. From the rate of disintegration of the newly formed radioactive element the number of neutrons absorbed can be calculated. By use of the capture cross section of the element of which the foil is composed the density of the neutrons can then be computed. By counting the alpha particles from a bismuth foil of known weight after exposure to neutrons the density of the neutrons which produced this activity can be calculated. The least accurately known value which enters into this evaluation is the capture cross-section of bismuth for pile neutrons. This value may be improved by future measurements. It is of importance only for absolute measurements of neutron density so that relative values can be measured with considerable accuracy on the basis of known data.