Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized in previous thermal analyses of these cores. Testing of the bottom-orificed SM-1 stationary element and the SM-1, SM-1A, and PM-2A control rod assemblies showed flow distribution exceeded plus or minus 12% devation from average. Simple modifications to the SM-1 stationary element indicated the possibility of improvng fiow distribution in that element. (auth)

Editor please delete 26456.<><DSN>16:026457<ABS>Work done in the P122 reactor control actuator area is summarized. Actuators were required to radially position the absorber blades in the core of the reactor. The P122C1 was a subsonic power plant and temperatures were low enough to permit the use of hydraulics in the actuator area. The program was reoriented and the power plant designated P122C3 which was a supersonic version of the folded flow power plant. The ambient temperature at maximum power was high enough to require pneumatic actuation of the control blades. The program was reoriented after two design iterations of the subsonic power plant. A test model of the actuating equipment and the entire linkage assembly was on hand and completed when the program was cancelled. The linkage was being redesigned for the supersonic application and special bearings were ordered for fabrication into the lower temperature rig. The actual mechanical concepts of the pneumatic actuator were under study when the program was cancelled. (auth)

The intent of this work was to explore more fully the pressing of sinterable UO2 powders into cylindrical compacts in the hope that a more precise prediction of green density in terms of powder properties, pressure, and geometry could be evolved.

Since it is now possible to obtain a heat-treated U core that is randomly oriented and has a finer average grain size, it is necessary to irradiate measured, transformed fuel cores over the full range of grain sizes, in order to compare relative dimensional stabilities. An improved ultrasonic tester, Fuel Core Tester-UT-2, is used to test all fuel cores.

Fuel element misalignment is an apparent cause of ``hot-spot`` ruptures. Several methods of eliminating hot spots have been tried, however, none appear to have completely solved the hot-spot problem. Attachment of projections to the side of the fuel elements appears to offer a means of minimizing misalignment since they act as bumpers against the side of the process tubes. Preliminary data from tests now in progress or recently completed indicate excessive corrosion rates are not to be expected on fuel elements with projection attached. In fact, no hot spots were found on 39 columns of normal self-supported I&E fuel elements discharged at normal goal or on four columns of enriched self-supported I&E elements irradiated to 850 mwd/t exposure.

At the request of IPD Personnel, calculations on neutron and gamma attenuation were made for the NPR shield. The calculations were made using a new shielding computer code developed for the IBM 7090. The calculations show the thermal neutron flux, total neutron dose rate, and gamma dose rate distribution through the entire shield assembly. The calculations show that the side and top primary shield design is adequate to reduce the radiation level below design tolerances. The radiation leakage through the front shield was higher than the design tolerances. Two alternate biological shield materials were studied for use on the front face. These two materials were iron serpentine concrete mixtures with densities of 245 lb/ft{sup 3} and 265 lb/ft{sup 3} (designated by I-S-245-P and I-S-265-P, respectively). Both of these concretes reduced the radiation below design tolerances. It is recommended that the present front face biological shield be changed from I-S-220-P to I-S-245-P. With this change the NPR shield is adequate according to these calculations. The calculations reported here do not include leakage through penetration in the shield.

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Current NPR fuel production plans call for a thick layer (0.030 inch or greater) of 5% Be + 95% Zry-2 braze alloy in the closure region. This requirement was imposed to eliminate many of the production welding problems brought about by the presence of the low melting braze alloy between two surfaces of Zircaloy. All GEH-4 irradiations in the past have involved very thin braze lines (0.015 inch or less). As a part of the fuel evaluation program it is essential to run a comparative irradiation to determine what effect the braze line thickness has on the stability of the fuel closure. For this purpose three fuel elements were prepared, two with a braze thickness of 0.030 inch and one with a braze thickness of 0.015 inch. To provide a more severe thermal stress, the I&E geometry was used. Five MTR cycles should be sufficient to test this fuel concept. This report details this test proposal.

This report provides data for the proposed HAPO H-1 Loop irradiation testing.

Thorough examination of the clad surfaces and end closures of the two brazed elements gave no evidence of exposed uranium. This examination included visual scrutiny in the Radiometallurgy Examination Facility, followed by dissection and metallographic study of suspected zones. Thus, it is concluded that the emission of radioactive material was probably due to a jacket or closure leak in one or more of the other experimental elements, included in the same tube, whose investigation is beyond.the scope of the present study. Metallographic examination revealed voids and unbonded areas in the closure region. Also, erosion and clad thinning was observed where the braze blended into the fusion weld. As a result of these observations, corrective action was taken to improve the quality of the brazed closure.

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This report discusses the request of the Hanford Operations Office, Atomic Energy Commission, for replacing the code word {open_quotes}Birch{close_quotes} with a new code word {open_quotes}Brandy,{close_quotes} and the word {open_quotes}Wine{close_quotes} to replace the code word {open_quotes}Bark{close_quotes} has been officially approved by the AEC Headquarters, Washington, DC.

A test on the Shippingport PWR was described, whose purposes were to determine if sufficient reactivity was present in the core to override a peak xenon transient, and to obtain data for rod worth calculations. Equilibrium xenon was established at 60 Mw net electrical output for approximately 35 hours prior to shutdown. The flux was then reduced to approximately 100 mu mu a. The reactor was maintained critical by withdrawing or inserting rods as necessary. Startup rates were measured by a mu mu ammeter. The xenon transient was followed in this manner for 60 hours. Override of peak xenon occurred 8 hr 50 min after shutdown. (auth)

This report addresses the design properties for graphite.

The detection of radioactive air filter fires is discussed. Criteria requirements for a suitable fire detection system were established. The applicability of aircraft-type fire detection systems for this use was evaluated. The operation of a discrete eutectic salt type continuous fire detection tubing system is outlined. (M.C.G.)

References to the data and a brief resume of the methods used in generating cross sections stored on magnetic tapes in fine energy detail are given. The data are used directly in bare, homogeneous reactor slowing-down calculations or can be processed to provide multi-group of multi-level cross sections in coarser energy detail for other machine programs. Cross sections for 33 elements or isotopes, the U/35 prompt and total fission spectra, and the Be(n,2n) transfer cross sections are currently available. (auth)

The rapidly increasing volume of new data in scientific and technical fields demands faster and better ways to communicate the new information to those concerned. A working system, System of Automatic Processing and Indexing of Reports (SAPIR), for doing this is described. SAPIR makes use of the Keyword-in- Context Index principle, by which certain keywords, together with surrounding words that act as modifiers, are selected from the titles of the technical publications. These keywords with their modifiers serve as index entries from each title, therefore, there will be as many index entries generated as there are keywords contained in the title. The index entries are sorted alphabetically by the keywords. The first letters of the keywords in each of the index entries form a column in the printed format, which makes scanning for the soughtafter item easier. The SAPIR system is automatic and the list of citations is machine- generated on an IBM 1401 Computer. (auth)

Activities are discussed for work done on the design, components development, and engineering analysis of the MSRE, and materials development studies including metallurgy, in-pile tests, chemistry, engineering research, and fuel processing. (B.O.G.)

Although there were the ususl fluctuations in background at certain of the monitoring stations, the contamination levels recorded do not differ significantly from those of the previous year except that there appears to be a slight trend downward. Data are tabulated. This downward trend can be attributed to a curtailment of operations at the Laboratory, the gradual implementation of the containment program, and a curtailment in world-wide weapons testing. Tall-out data are included. Two personnel exposures were recorded which have been reported elsewhere. One emPloyee received a relatively high exposure to the left hand which consisted primarily of soft radiation. A second employee apparently has accumulated a sizeable fraction of a permissible body burden of Pu/sup 239/. The number of unusual occurrences increased over the previous year. However, in general, these events posed only routine problems and it is probable that the noted increase in such occurrences is due primarily to a more complete reporting system which was inaugurated early in the year. (auth)

Reaction between gaseous nitrogen dioxide and gaseous uranium hexafluoride results in the formation of the nitrylium salt NO/sub 2/UF/sub 6/. Under similar conditions no reaction occurs between nitrogen dioxide and molybdenum and tungsten hexafluorides. Reaction between nitrosyl and nitryl fluorides and the three hexafluorides results in solid compounds of the composition NO/sub x/F x MF/sub 6/, where x is 1 or 2 and M represents Mo, W, or U. Reaction between gaseous or liquid nitrosyl chloride and molybdenum and uranium hexafluorides results in the formation of the nitrosylium salts NOMoF/sub 6/ and NOUF/sub 6/ and chlorine. Tungsten hexafluoride was found not to react with nitrosyl chloride. The conditions and stoichiometry of the reactions are described. The nitrylium salt, NO/sub 2/UF/sub 6/, is characterized, and some of its physical and chemical properties are described. The NO/sub x/F x MF/sub 6/ compounds are characterized, and some physical properties are described. Evidence for the presence of NO/sup +/ and MF/sup -//sub 7/ ions in these compounds is given. The nitrosylium salts, NOMoF/sub 6/ and NOUF/sub 6/, were characterized previously. (auth)

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

Because of various evaluation studies on fuel stability, it is proposed that Coors Porcelain Company change their recipe for fueled tubes. Specifically, zirconyl nitrate should be used. in addition to uranyl nitrate. The detailed changes to be made are provided.

Report discussing variations in comparative tests of portland cements that could lead to the rejection of a material fully conforming to specification requirements, or the acceptance of a material with undesirable chemical or physical properties. Many of the causes for variation in chemical analyses and physical test results are listed in this discussion, and remedies for some of the more frequently encountered deficiencies in apparatus and methods are suggested.