Abstract. A program was instituted to study and reproduce the in-reactor intergranular failures of Type-304 stainless steel fuel cladding found in superheated steam. The program was directed toward finding ways to eliminate the cause of failure or to use improved alloys that would be less susceptible to failure. A materials screening test was developed in the out-of-pile superheat facilities with 1.5 ppm chloride added as sodium chloride to the recirculating water in the presence of typical boiling water reactor quantities of oxygen and hydrogen. During the test, the heater sheaths were exposed through several cycles to saturated steam (with its accompanying moisture carryover) and superheated steam. Failure of Type-304 stainless steel was obtained in periods of less than two weeks; the failures were predominantly transgranular. Type-347 and vacuum-melted Type-304 stainless steels failed in this NaCl-cycle test while Inconel-600, Incoloy-800, Hastelloy-X, Type-406 stainless steel, and vacuum-melted Type-310 stainless steel were acceptable. An improved chloride cycle test with 0.5 ppm chloride added as ferric chloride to the recirculating water was developed. An intergranular failure was obtained similar to that experienced in the superheat fuel cladding failures in the superheat in-pile loops in the Vallecitos Boiling-Water Reactor. Sensitized Type-304 and Type-316 stainless steels …

Abstract: The question of how crack-free, protective oxide films can form on zirconium during oxidation when the Pilling-Bedworth ratio is about 1.5 has been considered by a study of the relative plasticity of various forms of zirconia. Hot hardness measurements showed that doping mono-clinic zirconia with iron, nickel, or chromium resulted in softer (more plastic) structures and that yttrium additions slightly reduced the plasticity. Calcia-stabilized cubic zirconia was found to be more plastic than mono-clinic zirconia when tested at temperatures above 200 degrees C. The behavior of anion-deficient oxides indicated that they were more plastic than stoichiometric oxides even though the hardness values were identical at 23 degrees C. The former were free from cracks at the indentions, whereas, stoichiometric oxides exhibited extensive cracking around and between indentions. The behavior of actual, thick (72 microns) oxide films during tensile deformation of oxidized metal samples indicated that considerable plasticity occurs in the oxide at 500 degrees C but that the films are brittle at 23 degrees C. It was concluded that the plasticity of the oxide may be greater than that of the oxygen-contaminated substrate at elevated temperatures and may be the means by which epitaxial strains are minimized.

A research program is being conducted to obtain experimental data in the irradiation of plutonium-enriched fuel to confirm a theoretical model for predicting isotopic composition and reactivity changes in plutonium-enriched, light-water-moderated reactors. Quarterly progress: Forty-six fuel pellet faces have been auto-radiographed. These faces have been prepared from twenty-three pellets by making an exposure before and after the removal of an additional ten mils of fuel. A substantial number of large "hot spots" continue to appear. The largest spot so far observed was 44 mils long, 20 mils wide, and of the order of 20 mils thick. This spot has a PuO2 concentration which varied from 70% on the periphery to 100% at then center. There is some evidence that the segregated regions are elongated with their long axes perpendicular to the direction of the pressing of the green pellet. Determination of the size and concentration distribution is continuing. The EPITHERMOS code now seems to be operating correctly. A test problem for a typical water lattice converged in eleven iterations. The computation of the spectrum for a pure water medium gave results which agreed very well with the expected 1/E spectrum. At the end of the quarter, the program fuel element …

Summary: All experimental work under the Corrosion Mechanism task has been completed. The remaining topical reports are being prepared by D. L. Douglass, now on assignment at Mol. Experimental work on the first round of 31 alloys and on the second round of 10 alloys has been completed. Steam exposures of at least 3000 hours were finished for all the alloys at all test temperatures, with exposures of some coupons to 6700 hours. Mathematical expression have been derived to describe all first round data for corrosion rates and hydriding rates at 300, 400, and 500 degrees C as a function of Nb, Cr, Fe, and Cu content. Solution of the equations for particular service temperatures yield Zr-Cr alloys at optimum at lower temperatures and Zr-Cu-Fe alloys as optimum at the higher temperatures. The second round test results show that neither Ni nor Be additions to Zr-Cr or Zr-Cu improve the performance over that of the optimum Zr-Cr or Zr-Cu-Fe alloys. For the first round heat treatment used, post-corrosion ductility depends on two factors in addition to alloy composition and hydrogen content: crystallographic texture and intermetallic aging reactions. Alloys with a high original ductility are embrittled less by a given amount …

The 300 series stainless steels had been selected originally as the reference fuel cladding material for utilization in several superheat reactor (SHR) systems being built as part of the United States Atomic Energy Commission (USAEC) program. The adequacy of the general corrosion resistance of Type-304 stainless steel for superheat fuel cladding was confirmed in the Phase I portion of the study reported previously. Fuel jacket failures that occurred in Type-304 stainless clad fuel elements exposed in the Vallecitos boiling water reactor superheated steam loop indicated the questionable dependability of such stainless steels for this SHX fuel cladding application. The following conclusions are based on the out-of-pile general corrosion evaluations completed to date on Incoloy-800 as a fuel cladding for nuclear superheat applications: 1. The corrosion data from 4000-hour heat transfer tests indicate good corrosion resistance up to at least a 1300°F metal temperature. By use of a devised method of data treatment, the general corrosion for three years exposure at 1300°F can be calculated to average 0.0016 inch with an upper 95 percent confidence limit of 0.0033 inch. 2. A compositionally-disturbed layer develops at the metal-oxide interface. The disturbed layer depth is a function of time and temperature of exposure. …

The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.

A research program is being conducted to obtain experimental data in the irradiation of plutonium-enriched fuel to confirm a theoretical model for predicting isotopic composition and reactivity changes in plutonium-enriched, light-water-moderated reactors. Quarterly progress: Work has begun in the Radioactive Materials Laboratory to sample the project fuel from the pins irradiated to 1800 and 5000 MWT/T. Some delay has been experienced due to preemption of the hot cells by priority work. Examination of the autoradiographs of the un-irradiated project fuel showed that in a volume of fuel approximately equivalent to a pellet there were 13 hot spots larger than 15 mils. Evaluation of these spots with the fuel analyzer showed that they contained about 14 mg of PuO2 or about 9% of the total present. The EPITHERMOS code is being modified to automatically normalize the epithermal scattering to the correct value for all moderators. Calibration of the flux wires has been made and the reduction of the data from the VBWR irradiation is nearly complete. A similar resonance activation was made in the water reflector of the Stanford Pool Reactor to obtain the relative activity in a well-defined pure water spectrum. Reduction of these data is also in progress.

Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. In the course of this program, a new theory was suggested and an experimental apparatus was designed and built. Experiments were carried out to test the new model. This present report contains additional data and information extracted from the experiments at PG&E Humboldt Bay Power Reactor at Eureka, California. During the last days of 1963 a number of control rod and fuel bundle worth measurements were made in the ESADA Vallecitos Experimental Superheat Reactor (EVESR) using the (k[beta]/[script l] technique. A description of the experiments is given in the text of the report and some results are reported. A computer program was written to perform the data analysis of the pulsed neutron experiments and the code is discussed in the Appendix.

Technical report describing an analytical investigation made on the mechanism of the "downstream" effect in the corrosion of stainless steel in sodium. A mechanism of iron alloy corrosion is assumed in which the controlling rate is diffusion of iron-oxygen species, probably a FeO-Na2O complex. A mathematical model is developed for the corrosion, and the predicted results agree with the experimental data. The corroding species is probably present in sodium at concentrations of ~10(-8) g Fe/g Na.

Summary: Stainless steel clad nuclear fuel cycle costs can be reduced to those associated with Zircaloy clad fuel or potentially lower by reducing the thickness of the clad tube wall until performance penalties offset the savings associated with the reduction in parasitic neutron absorption. To demonstrate the feasibility and investigate performance capabilities of thin clad fuel rods for power reactor application an assembly was fabricated with 0.0127 cm (5 mil) thick stainless steel cladding tubes for irradiation testing in the Vallecitos Boiling Water Reactor (VBWR). The fuel bundle was placed in the VBWR and irradiation was begun in November, 1961. The irradiation is scheduled to continue until the target exposure of 2.74 x 10(20) fissions/cc (10,000 MWD/T of uranium) average burnup is reached. Destructive examinations of fuel rods will be performed at regular intervals throughout life to determine fuel rod performance.

Results of a questionnaire survey are presented on reactor supervisor and operator education, experience, and training, and on certification procedures by reactor category. The questionnaires were obtained from 362 supervisors, 393 operators, 44 supervisor trainees, and 139 operator trainees.

From introductory paragraphs: "The central problems in neutron research are the understanding of nuclear structure and the study of the properties of nuclear particles, particularly the properties of the neutron. The most fruitful attack on these problems is the determination of the probability of interactions between neutrons and nuclei, i.e., the measurement of neutron cross sections. Ideally, this involves the study of all possible types of neutron interaction with all available nuclei at all neutron energies...The discussion in this paper will omit the interactions leading to neutron productions, and will be limited to the intersections of neutrons with stable nuclei."

Nuclear excavation is the name given to the concept of using large scale nuclear explosion craters for useful projects, such as harbors, canals, and roadway cuts. It is one of the principal applications of the Plowshare Program for industrial, or peaceful, uses of nuclear explosives. Plowshare is sponsored by the U.S. Atomic Energy Commission and is under the technical direction of the Lawrence Radiation Laboratory at Livermore, California. The purpose of this paper is to describe cratering concepts and the present state of nuclear excavation technology. The general nature of the safety hazards associated with nuclear excavation are also discussed.

Recent underground nuclear tests conducted by the U.S. Atomic Energy Commission have yielded data on the effects of contained nuclear explosions in four rock mediums: tuff, alluvium, rock salt, and granite. This report presents and compares data obtained primarily through exploratory mining and drilling into the postshot environment of 35 such events.

Late in 1962, a group in the Computation Division at Lawrence Radiation Laboratory, Livermore, began a study of compiler languages and techniques, the culmination of which was a machine independent FORTRAN written in FORTRAN. The impetus behind this study was a local need to move rapidly and efficiently from one machine to another. A secondary incentive was the need to be able to implement language extensions without the customarily long gestation period. Because of a large inventory of FORTRAN codes at this installation, FORTRAN source language was chosen as standard. Some effort was also expended on a syntax directed compiler written in FORTRAN for FORTRAN. With the knowledge of FORTRAN techniques gained from writing the syntax compiler and translators for other machines, the writing of FORTRAN in FORTRAN was begun.

In any underground nuclear explosion, the shock front that propagates from the shot point carries with it energy from the explosion, and distributes this energy by doing work on the surrounding material. In the process, the material undergoes changes in both its physical and mechanical states. If enough energy is deposited in the material, it will vaporize or melt thus changing its physical state, or cause it to crush or crack. During the past few years, special computer codes have been developed for predicting the close-in phenomena of underground nuclear explosions using the laws of physics, and the knowledge of the properties of the materials in which the detonations occur. As a consequence, a better understanding of experimental observations and measurements has evolved.

The effects of explosion-induced ground motion must be evaluated in planning and executing any nuclear excavation project. For some projects ground use intensity may dictate the use of less-than-optimum yields to minimize damaging effects. In remote areas, weighing the alternatives of outright purchase of some property or use of smaller yields may be required. The cost of indemnifying owners against damage must be considered in any case. Discussions of the effects of ground motion on three broad types of facilities - engineered structures, residential buildings, and equipment required for the support of nuclear excavation operations - are presented. A method of predicting the response of single- and multi-storied buildings, the response spectrum technique, is discussed, with emphasis on the application of explosion-induced spectra.

LRL has the responsibility of demonstrating the feasibility of a reactor for use as a power plant for a low-altitude, high-Mach-number missile. This reactor is literally a very high power air heater which must work at temperatures in excess of 2000' F. The reactor is exposed to high loads so one of the primary problems is providing high temperature structure. Considerable effort has been devoted to developing ceramic structural elements. One of the materials considered for this purpose is silicon nitride. In ceramic structural elements operating over large temperature ranges, a major problem is coping with thermal stress. In this respect there is a similarity with the radome problem. The work on silicon nitride at LRL consisted of limited fabrication studies (principally for familiarization), measurement of properties of interest to the application, and funding of fabrication scale-up efforts.

A fraction of the energy released by the underground detonation of nuclear explosives is locally deposited as residual thermal energy. An accurate prediction of this usable fraction of the energy released is necessary to evaluate the feasibility of several of the proposed projects in the Plowshare Program. This paper will present a summary of the available data on residual thermal energy from nuclear detonations in three different geological media: tuff, halite, and granodiorite.

The angle errors of a rotary table can be accurately measured by stepping off the angles with an optical caliper and computing table error from (1) the error readings at each angle measured and (2) the cumulative caliper error that will be evident when the circle is closed at 360', eliminating the necessity of adjusting the caliper to the exact setting.

To properly assess the "sensitivity" of explosives systems under impact conditions we must obtain detailed information on these mechanisms by which explosives are ignited by impact. It is necessary to know which impact conditions represent the greatest hazard to explosive materials, as well as to know for a given impact condition the relative responses of explosives of interest. I should like to describe to you a program sponsored jointly by the Department of Defense and the Atomic Energy Commission to attain this type of information. A wide variety of impact experiments have been conducted using geometrical arrangements of explosive, plastic, and metal which are relatively simple compared to complete ordnance systems. All of these tests have employed billets of explosive fabricated by standard production techniques, and the billets have been sufficiently large that ignition, once started, has something of the same opportunity to grow or decay as in full scale ordnance assemblies. From these relatively simple tests there has accrued much new useful information on the behavior of explosives under impact. With this new information providing a base from which to work, it is hoped that theoretical calculations on impact phenomena will in the future make a bigger contribution to …

Under the stimulus of electronic materials development - particularly thin-film studies - and the need for space environmental simulation chambers, a very rapid increase in the availability of industrial-sized vacuum components and systems operable in the ultrahigh vacuum range has taken place in the last three years. It is the purpose of this paper to explore the design considerations of ultrahigh vacuum systems for metallurgical applications.

The desired neutron spectrum for the High Flux Beam Reactor under construction at BNL requires use of U-Al fuel elements with more then 30 wt% U235 in the meat. The operating cycle of this reactor requires a minimum burn-up of 20%, and preferably 40% of the uranium in the element.

The purpose of this study was to investigate mathematically the concentration decrease due to particle deposition phenomena in highly concentrated monodispersed aerosols (mean particle size less than 1.0 mu) flowing through ventilation ducts. It was found that, from the standpoint of removal, the decrease in concentration due to deposition on duct walls was insignificant; but, when considering contamination on duct walls, the amount deposited, even though small when compared with the amount in the bulk stream, should not be overlooked.