Measurements obtained during three periods of HRT operations indicated that the xenon poison fraction was approximately 0.010. The technique used was based on mass spectrographic analyses of the stable xenon isotopes in the reactor off-gas stream. Models proposed to explain the measurements show that xenon, which is formed primarily by decay of iodine adsorbed on the pipe walls, is held up on the walls, out of the circulating stream, for an average period of about eight hours. (auth)

Molten fluoride salts, because of their radiation stability and ability to contain both Th and U, offer important advantages as high-temperature fuel solutions for nuclear reactors and as media suitable for nuclear fuel processing. Both applications have stimulated experimental and theoretical studies of the corrosion processes by which molten salt mixtures attack potential reactor materials. Corrosion experiments with fluoride salts which were conducted in support of the Molten-Salt Reactor E xperiment and analytical methods employed to interpret corrosion and masstransfer behavior in this reactor system are discussed. The products of corrosion of metals by fluoride melts are soluble in the molten salt; accordingly passivation is precluded and corrosion depends directly on the thermodynamic driving force of the corrosion reactions. Compatibility of the container metal and molten salt, therefore, demands the selection of salt constituents which are not appreciably reduced by useful structural alloys and the development of container materials whose components are in near thermodynamic equilibrium with the salt medium. Utilizing information gained in corrosion testing of commercial alloys and in fundamental interpretations of the corrosion process, an alloy development program was conducted to provide a high temperature container material that combined corrosion resistance with useful mechanical properties. The program …

The Molten Salt Reactor Experiment design, component development, and engineering analysis is discussed. Materials studies for the Molten-Salt Reactor Program including metallurgy, in-pile tests, chemistry, engineering research, and fuel processing are described. (M.C.G.)

The report comprises seven sections. A separate abstract was prepared for each section. (J.R.D.)

Thermoluminescence in potassium dihydrogen phosphate (KDP) induced by Co/ sup 60/ gamma irradiation at liquid nitrogen temperature (-198 deg C) was investigated. Glow curves in the temperature range --196 to 0 deg C were measured for a series of gamma exposure dosages ranging from 10/sup 4/ roentgen to 5 x 10/sup 6/ roentgen. The heating rate used for glow curve measurements was 12 deg C per minute. Twice recrystallized Mallinckrodt reagent grade potassium dihydrogen phosphate, with a grain size between 100 and 170 mesh, was used for most samples. In the case of the powder samples, the glow curve for an exposure dose of 10/sup 4/ roentgen exhibited two peaks in this temperature range, one at approximately --78 deg C and the other at approximately -146 deg C. The -78 deg C peak split into two distinct peaks with increasing dosage. At still higher doses an additional peak at about -9 deg C became evident. This peak may, however, be due to aluminum oxide. Calculation of the trap depth, E, and the frequency factor, s, associated with the --78 deg C peak, by the approximate method of Grossweiner yielded values of 0.485 ev and 9.97 x 10/sup 10/ sec/sup -1/ …

Tests of Cambridge absolute filters, Model Sl-071, specified for use in the GCR-ORR Loop No. 2 as full-flow, primary coolant fiiters were completed. kD.O.P/ (dioctylphthalate) efficiency tests were performed on three filters in the as-received condition, on two filters following canning and thermal cycling, and on one of the canned fiIters following bsking out. None of the three units met the design criteria of 99.97% efficiency for removal of 0.3 micron particles in the as-received condition. The postthermal cycle efficiencies of the canned fiIters were slightly higher than their respective as-received efficiencies. At the corapletion of testing, the two fiiters canned for installation in the reactor facility had measured efficiencies of 99.855% and 99.93%. These values were judged acceptable for the intended application/su The thermal cycling of the two canned filters and the subsequent baking out of one of these units demonstrated that a limited amount of off-gas products would be given off/su Pressure drop tests were performed on the canned fiiters with instrument air (ambient temperature, atmospheric pressure) over a flow rate range of 150 to 530 lb/hr. Curves of pressure drop across each fiIter versus Reynolds number were plotted for air and He. (auth)

Reported radioactivity in the Low-Intensity Test Reactor (LITR) water coolant traceable to uranium contamination on the surfaces of the alclad uranium-- aluminum plate-tyne fuel element led to an investigation to determine the sources of uranium contamination on the fuel plate surfaces. Two possible contributors to surface contamination are external sources such as rolling-mill equipment, the most obvious, and diffusion of uranium from the uranium-aluminum alloy fuel into the aluminum cladding. This diffusion is likely because of the 600 deg C heat treatments used in the conventional fabrication process. Uranium determinations based on neutron activation analysis of machined layers from fuel plate surfaces showed that rolling-mill equipment, contaminated with highly enriched uranium, was responsible for transferring as much as 180 ppm U to plate surfaces. By careful practice where cleanliness is emphasized, surface contamination can be reduced to 0.6 ppm U/sup 235/. The residue remaining on the plate surface may be accounted for by diffusion of uranium from the fuel alloy into and through the cladding of the fuel plate. Data obtained from preliminary diffusion studies permitted a good estimate to be made of the diffusion coefficient of uranium into aluminum at 600 deg C: 2.5 x 10/sup -8/ cm//sec. To …

The maximum stress calculated for the SMl-1 reactor vessel closure studs occurs during operation at full power. This value is 27,180 psi of which 19,800 psi is tension and 7380 psi bending. This stress does not include a stress concentration factor for effect of threads. It was eonservatively assumed the studs were initially tightened to a code allowable stress of 20,000 psi as specified in the ASME Code rather than the lesser stress obtained by the normal operating procedure. The maximum calculated stress occurs at the outside surface of the cover where the stress ranges from 318 psi in tension to 90,660 psi in compression. The alternating stress is 50,000 psi. According to the Navy Code for a stress range of 50,000 psi, the eover material ean safely undergo a maximum of 1600 cycles. It was estimated that the SM-1 will go through approximately 000 startup and shutdown cycles during a 20-yr life period, so the calculated stress is regarded as safe. For a transient eondition of 30 deg F/hr during heat-up, approximate temperature differences between the inside and outside surfaces of the cover were obtained. Temperature differentials between the inside and outside surfaces of the cover are increased by …

Dense partricles of ThO/sub 2/-UO/sub 2/ were prepared by a sol-gel process and vibratorily compacted into metal tubes to a density approaching 9.0 g/ cc. The steps in this method are all simple and can be carried out behind shielding, which is necessary for refabricating U/sup 233/ fuels. The sol-gel process consists of preparing a hydrous thoria sol, adding the U/sup 233/ as nitrate solution, evaporating to a gel, and finally calcining to almost theoretically dense oxide particles at orly 1150 deg C. The sol-gel-prepared oxide, after being sized, was compacted with a simple, inexpensive pneumatic vibrator. (auth)

Research was continued during the report period on the use of isotope neutron sources for producing short-lived radioisotopes. Experiments with a newly constructed betacounting cell are reported in which a 50-curie Be--Po neutron source was used. Study of the radiation chemistry of polymers was continued concerning the effects of polymer structure on free-radical formation. Free-radical formation in several additional polymers was studied. Preliminary work is also reported in an investigation of internal irradiation effects on the chemical activity of catalysts. (J.R.D.)

Science is impure in two ways. There is not a 'pure' science. By this I mean that physics impinges on astronomy, on the one hand, and chemistry on biology on the other. And not only does each support its neighbors but derives sustenance from them. The same can be said of chemistry. Biology is, perhaps, the example par excellence today of an 'impure' science. Beyond this, there is no 'pure' science itself divorced from human values. The importance of science to the humanities and the humanities to science in their complementary contribution to the variety of human life grows daily. The need for men familiar with both is imperative. We are faced today with a social decision resulting from our progress in molecular genetics at least equal to, and probably greater than, that required of us twenty years ago with the maturity of nuclear power.

The purpose of this document is primarily to discuss the effects of current retubing programs on the graphite moderator of the K Reactors. A secondary purpose of this report is to present general information related to graphite arrangement and keying patterns at the other operating reactors and to review, in a limited way, the types of graphite distortion observed at these reactors.

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Conclusions resulting from a recent corrosion study, together with evidence of graphite damage possibly occurring during retubing operations (having obvious implications in regard to reactor life), initiated a critical assessment of the water treatment process. This study was undertaken to determine whether immediate or near-term process changes could be made at the older reactors to minimize aluminum corrosion without causing adverse effects in other areas.

The increasing incidence of leaking process tubes, and the approaching end of the useful life of process tubes in the C and K Reactors, have focused attention upon the various sources of aluminum process tube leaks. Further, the replacement of large numbers of process tubes with new ones, also of aluminum, will require continued attention to these sources to make efficient use of the new tubes. One of the sources of process tube leaks is the corrosion that attacks the interior surface of the process tube. The factors influencing the extent of this corrosion attack are varied and complex, and in recent years, the corrosion service conditions have become increasingly more severe. Each of the factors involved in determining the rate of corrosion attack has thus become individually more important, and the need to understand the inter-relationships among them has increased. It is the purpose of this report to discuss the technical factors contributing to the internal corrosion of the process tubes, to review the way some of these factors have varied in the past, to examine the means available for evaluating the extent to which corrosion has damaged the tube walls, to comment upon the ways in which knowledge …

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The existing Hanford production reactors are enveloped by a graphite reflector, a cast iron thermal shield, and a biological shield consisting either of laminations of iron and masonite or of high density concrete. Cooling tubes were bonded into grooves in the cast iron thermal shield blocks with lead. As the graphite moderator continues to contract as a function of exposure, it is felt that the bond between the cooling tubes and the top cast iron blocks will be broken since these blocks are supported by the graphite. In the limit, the cooling tubes undoubtedly will pull from the grooves and be suspended between the cast iron and the top biological shield. In this event the effectiveness of the cooling system will be impaired severely, yet it will be necessary to limit the maximum temperature in (and the temperature gradient through) the top biological shield to assure its integrity. The calculations reported herein were made to provide heat generation rates in the shield complex as a function of fringe poison so that temperature distributions could be calculated for various postulated conditions of the cooling system. Thus, it should be possible to estimate the potential for controlling shield temperatures with fringe poison …

The criteria contained in this report have been established to provide the technical bases for the design modifications involved in the K-Reactor tube replacement. The ultimate intent of these criteria is to provide the basic technical data and concepts to assure: technical feasibility and operability of the reactor system as modified; operation of the reactor and its services to minimize nuclear and radiation hazards; and appropriate lifetime of the reactor and its service facilities as modified. The criteria are appropriately broad and may not contain all the data necessary to accomplish detail design. The information contained herein shall serve as a basis for evaluation and approval of all portions of the modification relating to the process as indicated in the above three points.

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Two simultaneous effects occur during the irradiation of polycrystnlline graphite which account for the observed dimensional instability. The first is a disordering of the crystallites which results in an expansion parallel to the c axis and a contraction parallel to the a axis. This process is highly temperature dependent and at room temperature is the principal effect. The second effect is a nonannealable contraction, which acts primarily in the transverse direction in anisotropic nuclear graphite, and seems to occur primarily in the non-graphitic part of the polycrystalline structure. (auth)

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A technique has been developed for slow cycle strain fatigue testing using specimens of thin-wall tubing of the type under consideration for use as super heat fuel cladding. Data on type 304 stainless steel have been obtained under radiation and in the absence of radiation. The strain cycle fatigue life of this alloy at 1200 to 1300 deg F is decreased three fold by the presence of a neutron flux of 1.5 x 10/sup 14/ > 1 Mev. Out-of-reactor data have been obtained on Inconel at 1300 deg F. (auth)

Fission product Ru, normally volatile to the extent of 20 to 60% in evaporation and calcination of simulated high level radioactive wastes at 500 to 1000 deg C, can be 99.9% retained in the solid product by addition of 2 moles of phosphite or hynophosphite per liter of waste. As little as 0.1 mole per liter lowered the Ru volatility during distillation approaching equilibrium by factors which varied from 38 for Darex to 225 for TBP-25 (aluminum) waste solutions. The final products with Purex, TBP-25, and Darex (stainless steel) wastes were insoluble glassy solids with densities of 2.4 to 3.8 g/ml and represented volume reductions of 2.9 to 8. These volume reductions are essentially the same as those obtained when the waste is calcined without additives. (auth)