Zircaloy-2 tubing was successfully joined to austenetic and ferritic stainless steel tubing by fusion bonding. Tungsten inert gas welding was employed to join a flange to a pressure tube. This work was performed to help determine the design of pressure tubes of a pressure tube reactor.

Existing experimental information about the nucleon-anitnucleon interaction is reviewed, and a description is given of a theoretical model, based on the Yukawa theory, which seems able to explain the experimental results.

Abstract: During March 1958, representatives of the health and Safety Laboratory, New York Operations Office, visited Columbia University to make stray radiation measurement at the Van de Graff accelerator. A radiation survey was performed at the Van de Graaff electrostatic generator at Columbia University. The survey results do not indicate a particular need for additional shielding at this time. It is recommended that the analyzer area of the beam room should not be occupied during beam operations and the present interlock system should be modified by the installation of a photocell interlock as shown on the included plan.

Output of plutonium from the separations plants Purex, Redox was slightly less than that scheduled; however, year-to-date production exceeds the corresponding commitment. Production of UO{sub 3} met the commitment, and UO{sub 3} shipments conformed to the established shipping schedule. Production of buttons and shapes was slightly less than the commitment as shown in the current Official Forecast. Shipments of buttons conformed to the current Forecast, while large shape shipments were slightly less than those forecasted. Feasibility of neptunium recovery was studied. Employee relations are reported.

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The use of organic cooling media for nuclear reactors operating at high power levels predicates the use of a coolant which will not react violently with metallic uranium in the event of a fuel element failure. This report describes the testing, and subsequent examination, of two pieces of irradiated uranium which were immersed in monoisopropylbiphenyl (MIPB) at high temperatures and pressures for periods of time up to twenty-five days. The uranium samples had different irradiation histories and cooling times. Similar experiments had been performed with unirradiated uranium by the Corrosion and Coatings Operation, and it was wished to determine whether irradiated uranium would react with MIPB in a different manner.

The purpose of this work has been to determine the production rate of tritium in fringe Li-Al alloy columns with the degree of precision necessary for economic analyses of such reactor loadings. These results are provided for use in such an analysis. This experiment indicates the production rate of tritium in the outermost fringe tubes to be T = 0.0216 M{sub E} = 0.175 M{sub t} where T = grams of tritium per full length (67 pieces) charge of Li-Al alloy material; M{sub E} = MWD/adjacent ton of E metal; M{sub t} = MWD/adjacent tube of E metal. The above values should apply for fringe loads utilizing greater or smaller quantities of E metal; that is, for isotope production loadings which are over or under-compensated from a reactivity standpoint. In the actual test load it was calculated that one gram of tritium and 13.5 grams of Pu were made for each 21.3 grams of U-235 burned up. During the same time interval the displaced uranium loading would have generated 24.3 grams of Pu and burned up 29.9 grams of U-235. The factor which seems to limit the accuracy with which these data can be interpreted is the ratio of the …

This is the monthly report for the Hanford Laboratories Operation. Metallurgy, reactor fuels, physics and instrumentation, reactor technology, chemistry, separation processes, biology, financial activities, employee relations, laboratories auxiliaries, radiation protection, operation research, inventions, visits, and personnel status are discussed. This report is for October 1958.

It has been noted on occasion that the side rupture rate at lower exposure is higher than one would expect in order to be consistent with the mathematical model (Weibull curve) used describe the rupture data. This has apparently been due to a few ruptures occurring at exposures lower than one would anticipate on the basis of the model. Restricting themselves to fuel elements irradiated under similar conditions of power and temperature, it is difficult to assert that these ruptures are anomalous with respect to the model used since there are so few ruptures involved. For this reason, some means of combining data was called for in order to see whether or not at least some of these low exposure ruptures were inconsistent with the model, and, if so, to revise the model in order to better describe side rupture performance. In this report, the results of an analysis of side failure data for failures covered in a previous report are given. Also included is a discussion of possible implications derived from recognition of the existence of this revised model.

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Brittle bonds and the factors controlling their formation have been of substantial concern in the production of dip canned fuel elements. Detection of brittle bonds has been by the chisel test and by metallographic examinations. At best, these are qualitative tests and do not establish the degree of brittleness. For this reason bond tensile strength analysis has been suggested. Tests have been run to determine if a change in canning variables could be detected by a change in bond strength.

Re-appraisal of the available data on flow times of the Columbia River between the reactor areas and Pasco was undertaken to permit extrapolation of the flow-time curves to lower river flow rates. Comparisons were made between data collected by the US Corps of Engineers and Regional Monitoring and with the equation for calculation of flow times developed by H.T. Norton. Extrapolation of the Regional Monitoring float study data to a flow of 3 {times} 10{sup 5} gallons per second was accomplished by comparison with the slope of the curve obtained from the US Corps of Engineers data; the latter covered flow times from 100-F Area to Pasco over a range of 3.4 {times} 10{sup 5} gps to 3.7 {times} 10{sup 6} gps. The revised flow-time curves are illustrated in Figures 1 through 6.

In the production of plutonium in a reactor, plutonium-238, 240, and 241 are formed as well as Pu{sup 239}. It is well known that the specific alpha activity of the plutonium varies as the percentages of these isotopes are changed. Kinderman, et al have worked out the relationship between isotopic content and MWD/ton exposure. Their findings are reported in this document.

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Attention has focused on the feasibility of conducting through-reactor, internal decontaminations in order to reduce rear face radiation levels and alleviate personnel exposure problems. The purpose of this report is to review the technical aspects of ``through-reactor`` decontaminations and to present an interim recommendation regarding the use of this method.

Twenty charges each of low hydrogen dingot fuel and standard ingot fuel elements were irradiated at DR-reactor. Distortion data from 22 tubes are analyzed in this document. The dingot fuel elements are prone to greater tube filling capacity (TFC) and greater daimeter growths at both center and ends of the fuel elements (the growths at center are larger than at ends). The difference between average warp values for the two types of fuel elements is not statistically significant, indicating that the greater TFC values shown by th dingot fuel elements are due to their larger diameter growths at the slug center rather than to larger warp values.

This report details initial results of a large scale accountability test program which was recently carried out in the Redox Facility. The test, as originally planned which was to consist of the complete processing (no inventory-clean plant basis) of about 55 tons of selected metal in conjunction with an extensive analytical, sampling, and volume measurement program. With the exception of two incidents, the processing requirements (minimum inventory and measurement of all material) necessary to the success of the test, were met. The two incidents which increase the uncertainties associated with some of the material balance values obtained were: the discharge of an estimated 700 pounds of uranium to the floor in a transfer from F-5 to F-4 due tot he improper installation of the F-5 to F-4 transfer line (jumper) and the discovery of a large accumulation of plutonium ({approximately} 15 kg) in the L-2 stripping tower after completion of the test run.

Four Zircaloy-2 jacketed enriched seven-rod cluster elements, one three-foot three-rod Zircaloy-2 jacketed enriched cluster element with integral end closures, and two eighteen-inch Zircaloy-2 jacketed wire wrapped cluster elements containing UO{sub 2} will be irradiated at jacket surface temperatures up to 270 C in the KER loops to an exposure of 3500 MWD/T on the enriched seven-rod cluster elements.

The Process Specifications for Operational Control of the Purex Plant(l) have been revised to define the operational control changes as a result of (a) converting the Purex Plant from three to two cycles, and (b) the use of the Plutonium Anion Exchange Unit (N-Cell) for routine plutonium concentration in place of the L-Cell Concentration Package. The flowsheets specified in Figures I-1, -2, and -3, and Table I fulfill the requirements for nuclear and chemical safety within the process as defined in Purex Process Specifications. The flowsheet also provides adequate recovery and decontamination to meet the product purity requirements itemized in Section 3.04. To clarify the two-cycle flowsheet and to simplify its use, the specifications are listed on graphic flow diagrams with the notation that process control allows a deviation of plus or minus five percent on all specifications except TBP concentration in solvent.

The objective of the test described in this report is to evaluate the effluent water contamination problems associated with the use of nickel plate corrosion barriers on X-8001 and C-64 clad I & E fuel elements.

Return can be increased on the order of $1000 to $2000 per calendar day at each K Reactor if the goal exposure for natural I & E slugs is reduced from 800 to 700 MWD/T. This gain is predicated on primarily a decrease in rupture losses. The recommended goal is a straight line on a linear graph of exposure in MWD/T versus maximum tube power in kw per tube, passing through the following two points: (850 kw/tube, 900 MWD/T); (1400 kw/tube; 490 MWD/T). In addition, I & E natural uranium charges running below 850 kw, if any, should be discharged not higher than 900 MWD/T. At a maximum tube power of 1300 kw, the recommended goal will give a weighted average exposure of about 700 MWD/T. Power at which a tube operates should be determined as recommended in HW-57937, ``Goal Exposure for Enriched I & E Material at the C and K Reactors.`` These are interim recommendations. We expect to recommend some changes in the goal, for reasons explained below. The recommended goal was calculated on the IBM-650, using an equation for return. The equation included, we believe, proper allowance for all but two factors. Principle affecting factors were included.

The objective of this test is to permit irradiation of projection fuel elements to evaluate jacket corrosion in areas of projection, to examine the effect of well-aligned fuel elements on hot-spot incidence, and to evaluate problems associated with projection slug usage in ribbed process tubes.