A thermal analysis was carried out to determine the temperature distribution in a SNAP-8 type reactor system during atmopsheric reentry. Of particular interest are the temperature distributions in the reactor upper head, the upper grid plate, and the vessel wall. The time and altitude were determined at which the upper head falls away from the reactor core due to having a portion of its wall melted through. The time and altitude of the melting of the upper grid plate and vessel wall were also determined. The effects of reentry attitude or equivalent angle of attack and initial temperature on the thermal behavior of the system were investigated. The computer programs used in various phases of the analysis were NEWTON (drag coefficient), RESTORE (reentry trajectory), and TAP (thermal model).

The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.

This report discusses the objectives of the FFTF Fuel Handling and Radioactive Maintenance System Group which are: To remove and replace test components from the reactor core under a controlled environment to permit maximum acquisition and preservation of experimental data; to service and maintain the reactor core to enable maximum utilization of the irradiation capability of the facility; and to provide the radioactive maintenance capability consistent with the requirement for high plant factor and safety considerations.

This paper shows the calculations for the Gas Volume Ratio, defined as the Total Gas Volume/Target Volume. Using 20 tons of LiAlO{sub 2} as the target, 5028 cubic feet of tritium, protium, and helium are produced. The target volume equals 312.5 cubic feet of aluminate, so that the GVR = 5028/312.5 = 16.1.

The ''handbook'' density for LiAlO/sub 2/ is 2.55 g/cc. Recent data show that it should be revised to 2.62 g/cc. In the future densities reported as percent of theoretical should be based on 2.62 g/cc as being 100 percent. Values quoted on this basis will be more realistic. In order to avoid confusion it will be necessary to state which number percent of theoretical density are based upon--that is, ''2.62 g/cc = 100 percent,'' or ''2.55 g/cc = 100 percent.'' For the convenience of those working directly with the material, a conversion chart is included along with a table which gives density (g/cc) versus percent density based on 2.62 g/cc.

The objectives of this test are to demonstrate the ability of a NERVA-type engine system to bootstrap from a tank pressure of 45 PSIA to full power, to study the stability of the control systems, to obtain high power mapping data and to demonstrate the stability of the core at high power. System response measurements will be conducted at both low and high power points to study the control systems. The operation of the engine at 4000{sup o}R chamber temperature will be mapped from design pressure down to the pressure at which the tie rod temperature limit is reached.

Isotope production calculations have been prepared in the past to indicate present and potential reactor capability; however, the cases evaluated generally considered production of only one or possibly two isotopes. With the wide variety and potential demand for numerous isotopes, the question arises as to what the production capability would be if several isotopes were being produced at the same time. To study this problem in more detail and determine which of several alternates may be used in meeting production requirements, the Case Generator and Evaluator System has been developed. This document presents a production matrix and reviews the assumptions used to derive the matrix. This matrix should be useful in planning future CAGE isotope cases.

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The attached pages show the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-1048.

Fire tests of bare, UF{sub 6}-filled shipping cylinders were conducted at the ORGDP Rifle Range during October 1965 as part of the AEC-ORO Container Test and Development Program presently under way at the ORGDP. The multi purpose effort was to determine if the cylinders would hydrostatically or explosively rupture; the time available for fire fighting before either incident occurred; and the degree of contamination as related to the type of UF{sub 6} release, wind velocity, and terrain. In addition to the cylinder fire tests, other tests were made for further evaluation of the fire-resistant BOX foam plastic. These included a newly designed shipping drum for 5-in.-diam cylinders, and 15B-type wood shipping boxes for small containers. In one case, the latter contained a UF{sub 6}-filled Harshaw cylinder. The test times ranged from 45 to 95 min. In no instance did temperatures exceed 200 F These tests are discussed under Part B. Our Nuclear Engineering Department was responsible for site preparation and the test program. The Safety and Health Physics Departments Mr. A. F. Becher, head, provided primary assistance in the conductance of the tests and was additionally responsible for the environmental monitoring and sampling. Personnel of the Plant Shift Operations and …

Douglas United Nuclear, Inc., has allocated to the Pacific Northwest Laboratory $542,000 of 02 Research and Development funds and $455,000 of Process Technology funds for FY-1967. Of these, $392,000 of 02 Research and Development funds and $420,000 of Process Technology funds are the responsibility of the Research and Engineering Section. The balance in each case is the responsibility of the Production Fuels Section. Representatives of these Sections have met with Pacific Northwest Laboratory personnel to develop programs to be undertaken in FY-1967. This document briefly summarizes the results of the discussions and delineates the work to be accomplished.

The pages of the report show the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-1048.

This document itemizes the irradiations performed by Irradiation testing for Pacific Northwest Laboratory. It lists the material undergoing radiation, awaiting disposition, and material shipped during the report period ending December 10, 1965. Materials irradiated were principally graphite but also included various steels and mixed fuel oxides.