5 S i ORNL-2517 ;72 4 C.85 - Rncm-hircm{r Nuclear Propulsion Systems (o AIRCRAFT NUCLEAR PROPULSION PROJECT ‘,\.J‘f { ¥ QUARTERLY PROGRESS REPORT G\ @’” FOR PERIOD ENDING MARCH 31, 1958 ¥ A A el OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the CENTRAL RESEARCH LIBRARY U.S. ATOMIC ENERGY COMMISSION DOCUMENT COLLECTION LIBRARY LOAN COPY DO NOT TRANSFER TO ANOTHER PERSON If you wish someone else to see this document, send in name with document and the library will arrange a loan. ORNL.-.2517 C-85 = Reactors-Aircraft Nuclear Propulsion Systems This document consists of 146 pages. Copy/y‘] of 219 copies. Series A, Contract No. W=7405-eng-26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT For Period Ending March 31, 1958 A, J. Miller, Project Coordinator DATE ISSUED AUG 201958 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tonnuuo operated UNION CARBIDE CORPORATION U.S. ATOMIC ENERGY COMMISSION MARTIN MARIETTA ENERGY SYSTEMS LIBRAR i il 3 4456 0251080 b C s FOREWORD The ORNL-ANP program provides research and development support in shielding, reactor materials, and reactor engineering to organizations engaged in the development of air-cooled and liquid-metal-cooled reactors for aircraft propulsion. The work described here is basic to or in direct support of investigations under way at General Electric Company, Aircraft Nuclear Propulsion Department, and Pratt & Whitney Aircraft Division, United Aircraft Corporation. This report is divided into four major parts: 1. Metallurgy, 2. Chemistry and Radiation Damage, 3. Engineering, and 4. Shielding. ;’ifi o ? INTRODUCTION The current ANP program at ORNL consists primarily of supporting research and developmental work for the General Electric Company, Aircraft Propulsion Department, on air-cooled and for Pratt & Whitney Aircraft on lithium-cooled reactors, Metallurgical, chemical, and engineering information required to develop and evaluate materials and components for use in specific systems is being obtained. The work in support of lithium-cooled reactor development is providing information on factors affecting the stability of high-temperature columbium-lithium systems, factors affectingthe mechanical properties of columbium, basic hydrodynamic and heat trans- fer characteristics of lithium systems, and com- ponents for use in lithium systems., Work under way on moderator materials for use at high temper- atures, shielding materials, and the production of low-oxygen-content yttrium metal is related to the requirements of both types of reactor. A comprehensive shielding program is being carried out to aid in the development of lightweight and crew shields for nuclear-powered Nuclear reactors reactor aircraft. The major experimental facilities em- ployed are the Lid Tank Shielding Facility (L TSF), which provides a beam of radiation for attenuation measurements on materials in slab geometry im- mersed in water; the Bulk Shielding Facility (BSF), which provides a reactor for attenuation measurements in three-dimensional geometry of configurations immersed in water; and the Tower Shielding Facility (TSF), at which operating re- actors, reactor shields, and crew shields can be suspended as high as 200 ft in the air, Much of the experimentation at these facilities is on material configurations and shield mockups speci- fied by GE or P&W in connection with their current design problems. Measurements are also made to obtain fundamental shielding data as part of a basic program of study of the theory of radiation attenuation and the development of shield design methods. Details of the progress made on this program during the past three months are presented in this report., Future progress reports will be on a semi- annual basis. ANP PROJECT QUARTERLY PROGRESS REPORT SUMMARY PART 1. METALLURGY 1.1. Fabrication Effort is being devoted to determining the effects of impurities, such as oxygen, nitrogen, hydrogen, and carbon, on the fabricability, weldability, mechanical properties, and corrosion resistance of columbium. In comparisons of cold-worked powder-metallurgy and arc-cast material, the form in"which the impurities are present was found to be as important as the quantity of the impurities, Studies of the effects of various heat treatments initiated, and the preliminary results indicated that nitrogen is the impurity most influential in heat treatment effects. Work was continued on the fabrication of co- lumbium tube shells. Three arc-cast billets were extruded satisfactorily and are to be drawn into tubing. Additional billets are being obtained for determining optimum extrusion temperatures and cladding procedures, Coatings of brazing alloys applied by flame spraying are being tested for oxidation protection of columbium. Both Coast Metals brazing alloys 52 and 53 were found to provide satisfactory protection in 1000-hr oxidation tests at 1700°F, Subsequent metallographic examination, however, revealed cracks in the coatings that could have led to rapid failure. Studies of the reaction rates between low-pressure gases and columbium were initiated on impure columbium were in order to determine the permissible limits of gaseous contaminants in the environment when columbium is exposed at high temperatures, Developmental work on methods and apparatus for the production of low-oxygen-content yttrium metal was continued. The method now being studied involves the lithium reduction of a low- melting mixture of the fluorides of yttrium, mag- nesium, and lithium to produce an yttrium-magnesium alloy from which yttrium sponge is produced by vacuum distillation. Castings prepared from the sponge by arc melting under an argon atmosphere with a tungsten electrode were analyzed for im- purity content, The chemical analyses indicated oxygen contfents af 600 to 3000 ppm and nitrogen contents of 50 to 17,000 ppm. Experiments are under way to determine the relationship between impurity content and tensile properties. An electron-beam method of consolidating the yttrium-magnesium alloy directly into a finished billet is being investigated. This method, if successful, could eliminate the vacuum-distillation and arc-melting steps. Also, equipment is being assembled for hydriding yttrium, 1.2. Corrosion Methods are being studied for purifying the lithium required in projects in support of both the air-cycle and the lithium-cooled reactors, Lithium as pure as it is practical to produce is needed for the reduction step in the production of low-oxygen- content yttrium metal, as well as for studies of the effects of minor amounts of oxide and nitride on the corrosion of reactor materials, Determi- nations of the solubility of nitrogen and oxygen in lithium and studies of various purification tech- niques indicated that gettering was a more satis- factory purification technique than low-temperature filtration or vacuum distillation, Initial gettering tests with zirconium, titanium, and a mixture of titanium and calcium indicated significant re- ductions of the nitrogen content of lithium, but these gettering materials had little effect on the oxygen content, Further experiments will be performed with other metals which have oxides whose free-energy values indicate that they are suitable as getters for oxygen. The extensive oxidation of an yttrium metal specimen upon ex- posure to lithium suggested the possibility of using yttrium metal in the purification of the lithium required in the reduction step of the yttrium production process. Two zirconium-base alloys being investigated for use as weld-filler rod in the fabrication of ductile saddle welds in columbium tubing were tested for corrosion resistance in lithium, Static and thermal- convection loop tests showed weld joints made with both an 85% Zr—-15% Cb and an 80% Zr-15% Cb-5% Mo alloy to have excellent corrosion resistance. Thermal-convection loop tests of lithium- columbium systems have thus far been of limited vii , Y duration because of the high rate of penetration of the longitudinal weld in the wall of the columbium tubing by the lithium metal, Two loops of sintered columbium tubing with inserts of arc-cast material showed very little attack away from the longitudinal weld, but there was severe attack in the heat- affected and weld zones. Static tests of lithium-filled columbium tubes were made in order to determine whether the en- vironment in which the thermal-convection loop tests were made had an effect on the weld attack. Four capsules, two in an argon atmosphere and the other two in a dynamic vacuum of less than 5 u, were attacked similarly. The weld and heat- affected zones of all four capsules were heavily attacked, and there were several cases of complete The base material away from the weld zone was unattacked. Experimental work on a lead-lithium alloy con- taining 0.69 wt % lithium - a potential shielding material — was completed. An experimental shield was prepared for studies at the Bulk Shield- ing Facility. penetration, 1.3. Welding and Brazing The effect of oxidation on brazed joints is being investigated. A nickel-rich solid-solution layer forms on the fillet surfaces of joints brazed with the Coast Metals nickel-base brazing alloys and the depleted region is quite ductile. Thus it appears that the ductility of oxidized joints of this type should improve during service. Studies of the rate of depletion of microcanstituents during high-temperature oxidation are under way. Refractory-metal brdzing alloys for use in lithium-cooled reactor applications are being in- vestigated, since the nickel-base brazing alloys have poor resistance to lithium, A survey was made of the columbium and zirconium binary systems of potential interest, and other systems will be studied. 1.4. Mechanical Properties Investigations were continued of the high- temperature mechanical properties of structural The testing programs include mechanical and thermal strain cycling, mechanical stress cycling (fatigue), creep in bending, and creep under multiaxial stresses. Inconel was chosen as the first test material because its tensile, metals. uniaxial creep, and relaxation characteristics are vili e 37 well known and could be used in the analysis of its behavior under more complex stress situations. In mechanical strain cycling, emphasis is placed on evaluating the influence of frequency of cycling on the number of cycles to failure. The extent to which the cycle time influences the number of cycles to rupture is dependent on the plastic strain per cycle and the temperature. A com- parison of the number of cycles to rupture at low strains for cycle times of 2 and 30 min re- vealed that long cycle times decrease the strain absorption capacity. This effect becomes more apparent at the low strain-per-cycle values ‘and is also magnified by increasing temperature, A severe frequency effect, for example, was observed at 1% strain and at 1500°F; fine-grained material endured approximately 700 cycles at 2 min per cycle and only 160 cycles at 30 min per cycle. At 1300°F, however, 2- and 30-min cycles yielded the same number of cycles to rupture at 1% strain, A series of thermal strain cycling tests at a mean temperature of 1500°F were completed at the University of Alabama. The correlation between mechanical and thermal cycling data was very Lack of agreement between the tests is attributed to the unreliability of the techniques presently used to measure the plastic strain per cycle in the thermal cycling tests, Fatigue studies were conducted by Battelle Memorial Institute, Data obtained at 1600°F on specimens stress cycled at 1 and 10 cps revealed a similar frequency effect to that observed in the strain cycling tests conducted at ORNL, Specimens cycled at the high frequency endured 10 times more cycles before failure than did the specimens in comparable lower frequency tests. The effect of cyclic stresses on creep properties was also investigated. Small alternating stresses were beneficial to the creep properties. The time to rupture showed a slight increase, and the creep rate decreased in comparison with pure creep properties, Upon further increasing the alternating stress, the time to rupture decreased. Conventional Goodman-type diagrams, which relate time to rupture to the stress ratio, were obtained for the two frequencies, 1 and 10 cps. Results from creep-bending tests at 1500°F the hypothesis that creep-bending be- havior can be predicted from tensile creep data, As postulated, creep in the outer beam fibers will follow curves in tensile creep tests at the poor, support stress initially present in the outer fibers. How- ever, after the first few hours, the bending creep rate correlates more closely with tensile creep data corresponding to a lower stress. This indi- cates that the bending stresses in the beam are redistributed through relaxation and that the re- distribution results in a lower stress in the outer fiber of the beams. Multiaxially stressed tubular specimens were creep tested at 1500°F. |t was found that the creep rate in the axial direction was not influenced by tangential stress in cases where the axial-to- tangential stress ratio, O’Z/Ort, was greater than 0.5. In other words, the axial stress alone de- termined the axial creep rate. At the stress ratio of 0.5, no axial creep was observed, although a tensile stress was present in the axial direction, At stress ratios, 0 /o, from 0.5 to -1.0, com- pressive axial creep was observed. Incompression, however, creep rates were accelerated by tan- gential stress. In contrast to the creep rates, time to rupture was greatly affected by tangential stress. As the stress ratio decreased from « to 1.0, for example, when one of the principal stresses was held at 6000 psi, the time to rupture decreased from 350 to 90 hr. At ratios below 1.0, where only the axial stress was varied, the variance of time to rupture with stress ratio was only slight, although the time to rupture continued to decrease as the stress ratio decreased, occurred at the stress ratio —1.0, The minimum time to rupture At present neither the creep nor rupture behavior can be related to the common mechanical theories of flow and fracture, However, a number of tenta- tive conclusions can be drawn from the data, It seems likely that deformation rates under multi- axial stresses are controlled by the maximum normal stress in the axial direction in the range of stress ratios from 0.5 to @, This implies that data from conventional uniaxial creep tests are valid bases for design under these conditions. However, the elongation to which a material may be stretched before failure is markedly affected by changes in stress distribution, and it decreases greatly with increased tangential stress. Recog- nition of this fact by the designer may alter the selection of a limiting strain. 1.5. Ceramics Dense beryllium oxide is being studied for use as a neutron reflector material in air-cycle The possibility of adding boron as a suppressor for secondary gamma rays is also being investigated. Specimens containing ZrB,, TaB,, TiB,, GrB,, HfB,, CeBG, and BN were prepared, and the specimens containing ZrB, and Hf82 were found to be much superior to the other combinations with respect to oxidation resistance at 1300°C. Additives are being sought that will densify extruded and cold-pressed BeO bodies and possibly, but not necessarily, fulfill the boron con- tent requirements, reactors, 1.6. Nondestructive Testing The study of eddy-current techniques for measur- ing cladding thicknesses was continued. Corre- lation of the measurements with the results of metallographic examinations is under way. Also, the possibility of measuring the thickness of a nonmagnetic cladding material on a magnetic core is being investigated. Measurements of aluminum coating thicknesses on steel in the absence of an intermetallic layer can be made with an accuracy of 0,0001 in. or greater, During investigation of the problem of measuring the thickness of aluminum coatings on steel, a new technique was developed which shows promise of eliminating the inaccuracies introduced by [ift- off or poor coupling between the probe coil and the part. A method that is independent of |ift-off variations will be adaptable to automatic scanning. An ultrasonic phenomenon Lamb waves, which are basically a characteristic elastic vibration established in a thin metal plate, is being investigated for possible use in the detection of laminations, The Lamb waves are generated in thin metal sections by directing a beam of ultra- sonic energy at particular incident angles to the sheet surface. known as Equipment has been obtained and is being tested for measurements of ultrasonic attenuation and velocity. Changes in ultrasonic attenuvation as a function of changes in metal structures will be examined, and methods for valid vultrasonic in- spection of welded structures will be developed. PART 2. CHEMISTRY AND RADIATION DAMAGE 2.1. Materials Chemistry The methods developed for the preparation of the LiF-MgF ,-YF, mixture required in the pro- duction of yttrium metal were analyzed in terms of possible large-scale production in available equipment. Since scale-up of the process appears to present no particular difficulties, a pilot plant in which the available equipment will be utilized was designed and is being constructed. The pilot plant will be capable of supplying a quantity of fluoride mixture sufficient for the production monthly of 200 Ib of low-oxygen-content yttrium metal. Experimental initiated of a possible method for the extraction of impurities studies were from lithium metal with molten salts. In order to avoid contamination of the lithium by extraneous metal, the eutectic mixture LiF-LiBr (12-88 mole %) was chosen as the extractant for preliminary studies. In an additional step, bismuth wasused to extract the residual lithium from the salt mixture following the initial impurity extraction. Further experiments mustbe performed before the effective- ness of the method can be determined. 2.2. Analytical Chemistry The precision of a method which has been pro- posed for the determination of lithium oxide in metallic lithium was evaluated by analyzing samples of a dispersion of metallic lithium in paraffin. In this method the lithium sample is dispersed in paraffin so that aliquots for analysis can be obtained by merely cutting and weighing. In order to determine the uniformity of the sample dispersion in the paraffin, dispersions were pre- pared with a Waring Blendor fitted with a stainless steel mixing cup in which a helium atmosphere was maintained. Samples of the solidified ma- terial, which contained approximately 1 g of the dispersed metal, were analyzed, The average value for five samples was 230 ppm, with a co- efficient of variation of 5%. These results indicate the validity of the assumption that lithium oxide is uniformly distributed in the lithium-paraffin dis- persion, | The apparatus previously used for determination of oxides in fused mixtures of fluoride salts is being adapted for use with BrF.SbF . This flux should be applicable to the determination of oxides in fused mixtures of lithium fluoride and magnesium fluoride. 2.3. Radiation Damage The tube-burst stress-rupture apparatus for tests under irradiation in the MTR was completed, and final preirradiation tests were initiated. The apparatus differs from that used in previous MTR irradiations in that the Inconel specimens will be irradiated in air rather than in helium to avoid changes in thermocouple calibration, and the specimens will be cooled internally in order to avoid overheating. While being irradiated the specimens will be at 1500°F and at stresses of 3000 to 6000 psi. It is planned to continue tests of Inconel specimens wuntil all experimental difficulties have been resolved and basic data on the material have been obtained. Metallographic examination of an Inconel stress- rupture specimen irradiated in the LITR for 600 hr while at a stress of 2000 psi and a temperature of 1500°F showed that rupture had occurred outside the gage length where the temperature was not monitored. The specimen was exposed to a fluoride fuel mixture on the inside and helium on the outside. No evidence of increased corrosion as a result of irradiation was found in the gage length where the temperature was monitored, A holding fixture for creep experiments in the ORR pool has been installed, and a flux monitoring facility was completed. Irradiation~induced changes in semiconductor diodes were studied in the light of the two princi- pal barrier theories: the Shockley diffusion theory and the charge recombination-generation theory, Experimental evidence indicates that irradiation causes a change in the predominant mechanism of rectification of germanium p-n junctions from the diffusion mechanism to the recombination-generation mechanism, This shift in mechanism explains the difficulty heretofore encountered in explaining the radiation-induced changes in current-voltage characteristics. Work continued on the preparation of apparatus for thermal-stress- and thermal-shock-resistance testing of moderator materials at high temperatures in the ETR. The specimens that will be irradiated include beryllium oxide, beryllium, and yttrium hydride. Preirradiation tests of the specimens will be made to determine elastic properties and dimensions and the effect of heat cycling. The yttrium hydride specimens will, in addition, be examined by x-ray diffraction in order to determine the crystal type and lattice spacing. These studies will be repeated after irradiation to detect changes. PART 3. ENGINEERING 3.1. Component Development and Testing loop was designed for studying the corrosion of columbium by lithium A forced-circulation metal circulating at high temperatures in a system with a temperature gradient. In order to protect the columbium tubing from air oxidation, the auxiliary portions of the loop, that is, the fill and drain lines, the line to the lithium expansion tank, and the loop which includes the rotating- magnet pump, will be clad with type 446 stainless steel. The tubing that will be at high temperatures in the test section will be surrounded with static sodium which will transfer heat to or remove heat from the columbium tubing in the hot and coid legs of the loop. The cladding of the external portion of the columbium loop will be joined with the stainless steel expansion tank of the sodium system. An all-stainless-steel loop of this design is being constructed in which NaK will be circu- lated in order to obtain operating data with which to validate assumptions made in heat transfer and flow calculations. The centrifugal pump rotary assembly in oper- ation under gamma irradiation in the canal at the MTR had received an accumulated gamma-ray dose of 1,6 x 107 rep by the end of the quarter, that is, about one-half the planned dose. The equipment continues to operate satisfactorily, and no gross changes have been observed in the oil {Gulfcrest- 34) which is used to lubricate the bearing and seal, Oil-lubricated pumps of this type are being tested for applicability to liquid-metal circulation in radiation fields, Modifications were made to the primary face seals of centrifugal pumps being tested for high- service. Use of a Fulton-Sylphon seal prevented accumulation of undesirable de- posits on the gas side of the seal. A split purge arrangement stopped the rise of vapor from the pumped fluid and likewise prevented accumulation of deposits on the seal face. One phase of the development and testing of shut-off valves for high-temperature service was completed. A valve was developed which is satisfactory for operation in NaK at temperatures vp to 1300°F, Cermet combinations known to be compatible with lithium were used as seat and The valve will be satisfactory if fabricated from columbium or another material temperature plug materials. compatible with lithium. 3.2. Heat Transfer Studies Experimental studies of the effect of thermal stress cycling on structural materials were con- tinued with the use of the pulse-pump system. The effects of exposure time, that is, total number of thermal cycles, on the Inconel test specimens are being investigated. For correlation with these studies, measurements are being made of the attenuation of sinusoidal temperature oscillations at the interior surface of a thick-walled Inconel cylinder, The study of the effect of screens in a diverging annular channel on the thermal structure of a fluid with internal heat generation was completed. The wuse of variable porosity screens reduced the amplitude of the surface temperature fluctu- ations to less than 10% of the axial rise at both the inner and outer channel walls. The screens also effectively reduced flow asymmetries caused by stoppage of one of the two pumps in the flow system, Initial studies of heat transfer in a liquid metal with internal heat generation were completed. Preliminary analyses of the data indicate that the surface-to-fluid mean temperature differences are greater than predicted. It thus appears that the theory-vs-experiment discrepancies observed in liquid-metal wall-heat-transfer systems — that is, externally heated systems — also exist in the in- ternally heated systems, 3.3. Instruments and Controls A data acquisition, or digital recording, system was obtained for use in recording and correlating the process variables for which measuring instru- ments will be required in a reactor system. The information obtained with this system will be use- ful in evaluations of the need for instruments and the preparation of specifications for instruments. Life tests of ORNL-designed liquid-metal-tevel transducers were continued., Two units have accumulated 5408 hr of operation at 1200°F in NaK and two have accumulated 775 hr. Eight units were tested to investigate wetting in NaK at various temperatures. The data are being analyzed. Two l-in. turbine flowmeters that had operated in NaK for 2000 hr at temperatures up to 1560°F were removed from the test loop and are being dis- assembled for examination. The design principles embodied in these instruments are suitable for the design of instruments for use with high-temperature lithium systems. Strain gages for static measurements at temper- atures up to 2000°F are being studied, Since past experience with strain gages has been almost entirely at low temperatures, a test facility for evaluating various types of strain gages at high temperatures is being devised, A design of an in-pile loop in which to study the effects of high-level radiation on thermocouples was prepared. The test environment will be sodium, Tests are under way on pneumatic pressure transmitters., Overrange pressures are being used to determine the aging characteristics of the instruments with respect to accuracy and over-all safety. PART 4. SHIELDING 4.1. Shielding Theory In order to predict and check the results of shielding experiments, a computation was made of the energy and angular distribution of the gamma rays at the surface of the BSF reactor. The calcu- lated gamma-ray energy spectrum normal to the surface of the reactor checked with that obtained in previous calculations, and the first detailed calculation of angular distribution was completed. 4,2. Lid Tank Shielding Facility The experiment designed by GE-ANP for studying the production of secondary gamma rays in con- figurations containing advanced shielding ma- terials is being continued. The latest tests were performed to determine the optimum placement of boral within configurations containing stainless steel, The most effective way of reducing the capture gamma-ray dose rates beyond the con- figurations was to distribute the boral evenly throughout the stainless steel and, in addition, to place some boral behind the stainless steel. 4.3. Bulk Shielding Facility The design of a stainless-steel-clad, U02- fueled reactor for the BSF was completed, and a report for submission to the Advisory Committee on Reactor Safeguards was prepared. Mechanical tests and analog computations carried out on the control system indicate that it is capable of safely re- versing positive periods as short as 8 msec. This would be equivalent to an excess reactivity of 1.2%; however, it is planned to limit the stain- less steel reactivity to 0.75% until a planned series of tests on the BSR-ll and its control system can be performed at the SPERT Facility. The study of the spectra of gamma radiation associated with the fission of U237 is continuing. It has included a second BSF determination of prompt radiation, the results of which are in general agreement with the previous results, although they differ by as much as a factor of 2 in some energy regions. Preliminary results were also obtained in measurements of the fission-product gamma radiation emitted during the period between 5 x 108 and 10~¢ sec after fission and during the period between 1 and 1500 sec after fission. A ?’/4-in.-diu by 2-in.-deep well was drilled into the truncated end of the 9-in..dia crystal of Nal(T1) — which will be used as a total absorption gamma-ray spectrometer at the BSF — in order to reduce the effects on the spectral response of the Compton scattering near the surface where the gamma radiation enters the crystal. In an attempt to improve the percentage resolution, seven 3-in.- dia photomultiplier tubes were mounted on the crystal, In spite of this improvement, the total spectrometer does not yet compete favorably with other spectrometers in terms of resolution, although its efficiency is two or three orders of magnitude greater. A study of the heat production in a large slab mockup of a typical aircraft shield was undertaken absorption at the BSF. The mockup being investigated con- sists of a 4-in,-thick slab of beryllium, followed by a gamma-ray shield and 16 in. of lithium hydride. The gamma-ray material consists of 3 in. of lead, 3 in. of iron, or 2 in. of Mallory 1000. The entire shield is submerged in an oil-filled aluminum tank positioned against the BSR. Measurements are being made of the heating, dose rates, and fluxes within the shield. 4.4. Tower Shielding Facility An experiment in cooperation with Convair, Fort Worth, is now being conducted at the Tower Shielding Facility to obtain neutron and gamma- ray dose rates and thermal-neutron fluxes under conditions similar to those encountered with the Nuclear Test Airplane Program but in the absence of the airplane structure. The experiment will also include dose rates and flux measurements taken at various positions around the ASTR while the reactor is 12,5 ft above the ground as in the Convair ground test measurements. In order to sort out the ground scattering effects, similar measurements will be made while the reactor is | suspended at an altitude of approximately 190 ft. Other measurements will be made to determine the spectra of neutrons emitted from the ASTR. xiii CONTENTS INTRODUGCTION ...ttt st ettt et ee e e e e e es e oo SUMMARY L.t e ............................................................................. PART 1. METALLURGY 1.1, FABRICATION ..o Columbium Investigations...........ccccoveviievceieiien e Fabrication Studies ..., Yttrium Metal Production ......cccoooiviiiiiiiiiiiie A0y Production ...ttt ettt eneea Evaluation of Material ... Yttrium Hydriding ..o 1.2, CORROSION ...ttt e, ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ Lithium Purification: Methods of Reducing the Nitrogen and Oxygen Content .........c.ooeovevvennnnn.. Solubility of Nitrogen and Oxygen in Lithium.... Filtration ..o Static Lithium Corrosion of Zirconium-Base Alloys ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ Thermal-Convection Loop Tests of Lithium in Columbium woooooeioeeeoe e Static Lithium Corrosion of Columbium .ooveeveiveinil 1.3. WELDING AND BRAZING .......cccociiriiiiiiieiienn, ............................................................................ ............................................................................ ............................................................................ Oxidation of Coast Metals Brazing Alloys 52 and 53 ..o, Development of Brazing Alloys for Lithium Service 1.4. MECHANICAL PROPERTIES ... Strain-Cycling Investigations.....c..cccveceivinieiciniecnen. Creep Under Bending Stresses..........cocveeeiivinceeennnn. Creep Under Multiaxial Stresses .........cccoovivineiinnnnn. 1.5. CERAMICS ..o, Boron Compounds ......cooveiiiiviieiiiiieees e Boron Additions to BeO ...occoooiviiiiiiciii Densification of BeO ... 1.6. NONDESTRUCTIVE TESTING ......ccoooviieee, Eddy-Current Thickness Measurements .................... Lamb Ultrasonic-Wave Studies ..o ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ ............................................................................ Xv 2.1 2.2. 2.3. 3.1. 3.2, 3.3. XVi PART 2. CHEMISTRY AND RADIATION DAMAGE MAT ERIALS CHEMIST RY oot ettt veeeeeerveassareesssesesrssseeatesessbsss s bbb eat e b s e an s b e sa e R e s e sat st et eatean et st sas s 53 Preparation of Charge Material for Reduction 10 YHIIUM ..cocriiciiii 53 Conversion of Y, 0,5 10 Y 4 oo 53 Preparation of MgF o oot 53 Preparation of the LiF-MgF ,-YF 3 Charge Mixture ..o, 54 The System LiF-YF 5 o 54 Extraction of Lithium Metal Impurities with Molten Salts ... 55 ANALYTICAL CHEMISTRY Lottt ot 57 Determination of Oxygen in Metallic Lithium ... 57 Determination of Oxygen in Fluoride Salts ... 57 RADIATION DAMAGE ...ttt ettt eb e eb s s b e saes e s encr e e e s e e st eas s abe e b r e s e et abeant e ste s 58 Creep and Stress Rupture Tests Under lrradiation ... 58 MT R EXPEFIMENTS .oiteiie et sr st s s e bbb bt 58 L TR EXPEIIMENES .oveiiiiiei ettt e n bbb b 59 DR R T @SS vnviereiereeeeeeeeee e eeeeaeassseaasesnsesaesaeseesssassntasstas st e abe e as s eme e saes s s e e e s s b s e e e R e e e e e A e e Renarr st et e et s 60 Radiation Effects in Electronic Components .......coevuiieeiciccnimiitieiieei it e e o 60 Irradiation Tests of Moderator Materials for Use at High Temperatures .......ccocovvvnivnniicnnn 62 PART 3. ENGINEERING COMPONENT DEVELOPMENT AND TESTING ..o cittsse sttt s vmne e 67 Loop for Investigating Corrosion in Lithium-Columbium Systems ..., 67 PUmpP DevelOpPmMEnt ...t s 67 Irradiation Test of Bearings and Seals ..o 67 Sl [MPrOVEMENTS ..oeieieececc ettt ettt e bbb s eSS 67 CAVIEAHON TESES ooiieieeiiieetietie it ii e et e veeeeeetees s eserteeabebserse et e s a e e aean s s eheeas e e ns s enees e e ateesbesasabbeab e e s b et r et 69 Valves for Use at High Temperatures .........co.ooviiiiiiiic e 69 Thermal Stability Tests of Metal Shells ..o 69 HEAT TRANSFER STUDIES ..ottt sttt s se s st ssa s e sn e sae b e san et asna s et st sonssnsne e 72 Studies of the Effect of Thermal-Stress Cycling on Structural Materials ... 72 PUlSE-PUMP SYSTEM ..oiiiieiiiiie it s 72 Thermal Entrance Studies with Water ... e 74 Thermal Structure of Fluids Flowing in Volume-Heated Systems ..o 77 Diverging Annular Channel with Screens ... 77 Liquid-Metal EXPeriment . .....cccoccicoiiiiiiiiiiiiiiiniice et 79 INSTRUMENTATION AND CONTROLS .ottt b s s 81 Data AcQUISTTION SYSTEM .o.eiviiiieieieere ettt e b b bbb 0 81 Liquid-Metal-Level Transducers ... 81 Strain Gages for Use at High Temperatures ... 81 Test Set-Up for Study of Effect of Radiation on Thermocouples ..o 82 Pressure Transmitters for Use at High Temperatures ... 82 4.1. 4.2. 4.3. 4.4, PART 4. SHIELDING SHIELDING THEORY ..ottt et s e st st et s ee e ee A Calculation of Energy Spectrum and Angular Distribution of Gamma Rays at Surface of BSF Re@Ctor ...ttt e ettt re e G BOMEIIY ...t e et e e et e e e e e e aneeeeanneeteteeebnneeeeaeeteeenntearnabeaeibeterbesatraranan Thermal-Neutron FIuX .. ...t e bae e s b e ra e e smenreesbeeene o GaMMA-RAY SOUPCES oottt et e et e e eba e s b aessan e be e bbaaearensane s aense o Absorption Coefficients ...t e st e et e re e BUIlUD FaCtOrS ..ot ettt e sttt ab b e sae e be s esbee s beeseeanaesrenn on Flux ComPutGEIOn ..cooeeiieeeeeeeeeee ettt eee et e et eb e b eab et e st e besateasbaaassaasasbeensensnteensesnnnsnanas Angular Distribution ...ttt CONCIUSTONS .ottt ettt et e st e e te e e aeete e s st e et e sebeesaesaseanseanseentesasennasasnesnn oo LID TANK SHIELDING FACILITY oottt sae e e Study of Advanced Shielding Materials (GE Series) ......cococoiviiinciie e BULK SHIELDING FACILITY ettt et ettt ekt eb ettt ebt b saeseeebene o The BSF Stainless Steel-UO, Reactor (B O R -} e ettt Spectra of Gamma Radiation Associated with the Fission of U237 . e, Fission-Product Gamma Rays Emitted More Than 1 sec After Fission ... Fission-Product Gamma Rays Emitted Between 5 x 1078 and 10~ ¢ sec AL Er S SI0M o ieitiiiiiteieeet et ettt e s et e e ce e et oot e e etsa e e e aaeee et aeaeemseeeenteeeaenneeeeaneeeeansee nraeeeebaeeeannbenaee e Prompt Gamma Rays ..ot ettt e et e et bb e b e e st et bt e b nn e esnen e r et s Total Absorption Gamma-Ray SpectroSCopy ......cooiviireiiiiiiicecrie e A Study of Radiation Heating in Shields ...........cooiiiiiiiiiiiiic e TOWER SHIELDING FACILITY ettt ettt ettt s b et s see e n s Aircraft Shield Test Reactor Experiment at the Tower Shielding Facility ... xvii Part 1 METALLURGY W. D. Manly Metallurgy Division T. Hikido U.S. Air Force 1.1. FABRICATION J. H. Coobs H. Inouye D. O. Hobson W. J. Werner Metallurgy Division T. Hikido, U.S. Air Force COLUMBIUM INVESTIGATIONS H. Inouye D. O. Hobson Fabrication Studies Cold Working. —~ Attempts have been made to fabricate columbium at room temperature because of its reactivity with gases at high temperatures. Metal prepared by the powder metallurgy process, in which relatively large quantities of oxides, nitrides, and carbides were present, was suc- cessfully fabricated, but arcecast material, in which the impurity content was lower, was brittle when fabricated at room temperature, It appeared therefore that the form in which the impurities were present in the metal was as important as the quantity present. The impurities are taken into solution in the melting step, and, if they do not precipitate significantly during cooling, a subsequent, suitable heat treatment of the material should improve its fabricability. Thus heat treatment of the metal at temperatures above the solubility limit of the impurities should have no effect on the fabricability, whereas, at a lower temperature, precipitation should occur and result in a ductile metal because of overaging. In order to test this hypothesis, sections of an arc-cast ingot prepared by the electron-beam melting process were heat treated at 750, 850, and 1000°C for 24 hr. The lower temperatures were selected to cause precipitation, whereas the higher temperature was selected to be above that at which the impurities would be retained in solution. The results of cold rolling and bend tests of the heat- treated specimens are presented in Table 1.1.1. Samples from the as-cast ingot and the sample annealed at 1000°C could not be rolled, but the other samples were cold rolled to foil. On the basis of the solubility data for oxygen! and nitrogen? in columbium, it appears that the element which causes the observed heat-treatment effects VA. U. Seybolt, J. Metals 6, 77476 (1954) 2C. Y. Ang and C. Wert, J. Metals 5, 1032—36 (1953). is nitrogen. The various specimens were identical, metallographically, after heat treatment. Extrusion. — Work has continued on the fab- rication of columbium tube shells. Three arc-cast billets were extruded satisfactorily and will be shipped to Superior Tube Company to be drawn into tubing. The three billets, all from Superior Tube Company, were copper plated for protection, heated in argon, and extruded at a 7:1 ratio. The first billet was extruded at 900°C, and, presumably as a result of contamination caused by pinholes in the copper plating, the extrusion had numerous small, deep cracks across the surface. The second billet was extruded at 815°C. The surface was fairly good, but there was a little roughness as a result of large grains and of some glass lubricant having been carried through the die unmelted. The extrusion pressure was approximately 60 tsi, that is, well within the limits of the press. The third billet was extruded at 980°C without the glass lubricant, and an extrusion pressure of only 53 tsi was required. The surface condition was very good. Toble 1.1.1, Effect of Heat Treotment on the Fabricability of Cast Columbium?* Heat treatment: 24 hr in vacuum of 10> mm Hg at indicated temperature Bend Angle Heat Treatment Cold-Rolling Temperature at Fracture** Results ©c) (deg} As cast 127 Brittle fracture an first pass 750 180 Rolled to foil 850 180 Rolled to foil 1000 Brittle fracture on first pass *Impurity analysis: C, 0,01%; 02, 0.12%; H,, 0.001%; N,, 0.032%. **Specimen dimensions: 0.125 x 0.340 < 1 in. ANP PROJECT PROGRESS REPORT Twelve columbium billets have been ordered for future work. These will be used to determine the optimum extrusion temperatures and the best cladding procedures. It may be possible by rapid heating in an induction furnace to extrude bare billets. Oxidation Protection Some work has continued on the oxidation protection of columbium. Satisfactory results were obtained in oxidation tests for 1000-hr periods at 1700°F of specimens coated with Coast Metals alloys 52 and 53. The specimens were coated by flame-spraying with the powdered alloy and were then heated in a purified helium atmosphere to a temperature just below the brazing temperature, A specimen coated with alloy 53 showed a weight increase of 1.1 mg/cm? in 200 hr, after which the weight decreased slowly to a total weight gain of 0.6 mg/cm? after 1000 hr. This weight change compares quite favorably with that which would occur in the same period if the specimen were type 316 stainless steel. At 1600°F, a type 316 stainless steel specimen would gain about 1.5 mg/em? in 1000 hr. A specimen coated with alloy 52 showed a weight increase during the first 24 hr and then lost weight for 100 hr, at which time the total weight change was a loss of 1.2 mg/cm2. After 300 hr the specimen began to gain weight again and it was still gaining after 1000 hr. Metallographic studies of the coatings showed a large amount of surface oxidation, some reaction loyer formation, and a few diffusion voids at the interface. The Coast Metals coatings were cracked in several places, but hardness readings showed no hardness increases in the columbium near the cracks. The specimen coated with alloy 53 had an average hardness of 110 DPH, while the speci- men coated with alloy 52 had a hardness of 78.7 DPH. The coatings ranged in hardness from 250 to 750 DPH. Oxidation tests were also run on two columbium tubes coated with alloy 53 and filled with lithium. The specimens were tested at 1500°F, and they failed because of incomplete protection by the The heat treatment evidently did not cause complete flow of the braze metal. brazed coating. Reaction Rates The reaction rates between low-pressure gases and columbium are being studied to determine the permissible limits of gaseous contaminants when columbium is exposed at high temperatures. An apparatus is being constructed that is designed for determining reaction rates of columbium with the pure gases at pressure levels of 0.01 to 1000 z Hg. In addition to providing a means for establishing the required purity of the atmosphere during high- temperature testing, the apparatus will be used for the preparation of samples for the determination of the effect of a specific impurity on the proper- ties of the metal. Preliminary absorption isotherms for oxygen in columbium at 950 and 1050°C show that columbium rapidly absorbs oxygen in the pressure range from 10 to 1000 i Hg. The absorption rate decreases as the pressure is decreased. In the pressure range between 40 and 800 ;. Hg, the formation of an oxide film competes with the absorption of oxygen by the metal. YTTRIUM METAL PRODUCTION T. Hikido W. J. Werner Alloy Production Six additional experiments were completed in the study of the lithium reduction of a mixture of the fluorides of yttrium, magnesium, and lithium to produce an yttrium-magnesium alloy. In run L-8 and subsequent runs both the lithium and the fluoride mixture were transferred in the molten state into the reduction retort through resistance- heated lines from their respective storage con- tainers by means of differential gas pressures. At the present time the transfer lines are con- structed of Z-in.-OD, 25-mil-wall Inconel tubing. The direct transfer of the molten fluoride mixture from the fluorinating apparatus to the reduction retort is now being considered. When this change is made a 3/8-in. transfer line will be used. This will facilitate the handling of larger quantities of material, and larger quantities of alloy will be produced in each run. A schematic diagram of the apparatus now being used is presented in Fig. 1.1.1. A distillation apparatus capable of handling 4 to 5 |Ib of alloy is used in conjunction with this apparatus. The ytrium sponge produced by distillation is arc-melted under an argon at- mosphere with a tungsten electrode to form finger PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED ORNL —LR-DWG 30443 —— - | - N . — TRANSFER TUBE ARGON OUT ~a— (z--wacuum AND ARGON IN THERMOCOUPLE TUBE — ] PRESSURE PRESSURE dll bPAE 4 : 7% TRANSFER TUBE —m \/ N \\ f F— NN /) Z % 7 % WATER JACKET — UL L Z z . LITHIUM POT SALT POT £ ) g ] % R THIN - WALLED TANTALUM CRUCIBLE % ST HEAVY - WALLED TANTALUM LINER GRAPHITE LINER Zz v 7 % Fig. 1.1.1. Diagram of Reduction Retort for the Preparation of Yttrium-Magnesium Alloy, (Secret with caption) castings of the type shown in Fig. 1.1.2. The results of chemical analyses of several of these castings are presented in Table 1.1.2. A new method for consolidating the alloy directly into a finished ingot is being investigated. The electron-beam melting method developed at Temescal Metallurgical Company is being con- sidered for this application. In the near future several ingots of yttrium-magnesium alloy will be sent to Temescal for melting by this process and will be returned for evaluation. If the process proves to be satisfactory for melting this alloy, the wvacuum distillation and arc-melting steps can be eliminated. Evaluation of Material Tensile Tests. — A series of specimens were machined from finger castings and pulled at room temperature. The purpose of these tests was to see whether there was any direct correlation between tensile properties and impurity content. However, as can be seen from Fig. 1.1.2, the grains in the as-cast specimens are quite large and nonuniform. A program for determining the recrystallization properties is being initiated in order to develop means for controlling the grain size. It is believed that, when the size of the grains of the tensile specimens can be con- trolled, a correlation between impurity content and tensile properties will be found. ANP PROJECT PROGRESS REPORT UNCL ASSIFIED Y-24851 040 in./ DIV Y 5 Y S Figs 1.1.2. Yttrium Finger Castings L-8A-1, L-8A-2, L-8A-3. (Secret with caption) Metallographic Techniques. — In the past the metallography of yttrium has been hampered by the lack of adequate polishing and etching technigues, but a suitable polishing and etching technique for yttrium has now been developed.® The specimen is ground dry on 320, 400, and SDevalopud by M. D. Allen of the Metallography Sec- fiflfll Table 1.1.2. Resulis of Chemical Analyses of Castings of Yttrium Sponge Run Oxygen Content Nitrogen Content Number (ppm) (ppm) L-8A-1 640 710 L-BA-2 1700 1,300 L-8A-3 1900 8,900 L-BA-4 3000 17,000 L-9A 1600 320 L-9B-1 1300 1,100 L-9B-2 1400 570 L-9B-3 1600 1,700 L-9B-4 1800 2,000 L-9B-5 1200 1,400 L-10A-1 610 140 L-10A-2 940 110 L-10A-4 1100 120 L-10A-6 1100 140 L-10A-7 980 100 L-10A-8 850 100 L-11A-1 1200 58 600 papers and then ground on 2.0 paper with a petroleum solvent and paraffin as a lubricant in the rough polishing step. It is then polished on a diomond wheel with 20- to 40-u diamond paste and canvas cloth, 6- to 8- diamond paste and canvas cloth, and 0 to 1-p¢ diamond paste and “‘airplane’” cloth. For the final polishing step, a Syntron vibratory polisher is used, with (1) Linde A on micro cloth and alcohol plus 10% oxalic acid as the lubricant, (2) Linde B on micro cloth and vacuum-diffusion-pump oil as the lubricant, and finally (3) magnesium oxide on micro cloth ond vacuum-diffusion-pump oil as the lubricant. The etchant used is an equal-volume mixture of HNOa, H,PO,, and CH,COOH. The etchant is cooled with ice in order to assure a slow etching reaction, and the sample is sub- merged in the cold etchant for 3 to 5 sec or until a thick film is formed on the surface of the specimen. The sample is then immediately washed with hot water. The reproducibility of this procedure is illustrated in Figs. 1.1.3, 1.1.4, PERIOD ENDING MARCH 31, 1958 Y-25111 w - L X o z . |o.03 Ee3 s O o Figs 1.1.3. Sample L-10A-2 Under Bright Field lllumination. 100X. Etchant: HNOB-H3F04-CH3CD'DH- | UNCLASSIFIED Y-25112 » Fy 0 - 1j— x Q = 008 I IE.'! > -o— e Figs. 1.1.4. Sample L-10A-2 Under Polarized Light lllumination. 100X. Etchant: HN03-H3PD4-CH3CUOH- ANP PROJECT PROGRESS REPORT and 1.1.5. All three figures are photomicrographs of the same area of sample No. L-10A-2 after 30% reduction by swaging and annealing at 800°C for 30 min. Figure 1.1.3 is a bright-field photo- micrograph; Fig. 1.1.4 is a polarized light photo- micrograph; and Fig. 1.1.5 is a sensitive tint photomicrograph. Yttrium Hydriding Several pieces of hydriding equipment were received from Wright Air Development Center. The equipment consists of two hydriding furnaces with associated gas metering tanks, a dissociation- pressure-measuring furnace, a modulus of elas- ticity rig, and a thermal stress rig. A program of study of hydrided yttrium is to be initiated. UNCLASSIFIED | Y-25113 | ; 1 - ~ ] -li.l"" =z Qo = 0.02 0.03 » - O o Fig. 1.1.5. Sample L-10A-2 Under Sensitive Tint lllumination. 100X. Etchant: HNOSHSPOA'CHECDDH‘ PERIOD ENDING MARCH 31, 1958 1.2. CORROSION E. E. Hoffman W. H. Cook D. H, Jansen Metallurgy Division LITHIUM PURIFICATION: METHODS OF REDUCING THE NITROGEN AND OXYGEN CONTENT Various means for purifying lithium, including low-temperature filtration, vacuum distillation, and gettering with active refractory metals, are being studied. Commercially produced lithium vsually contains oxygen and nitrogen in quantities of several hundred to several thousand parts per and techniques for reducing the total oxygen and nitrogen impurity content to less than 50 ppm are being sought. Lithium as pure as it is practical to produce is needed (1) for corrosion tests of container metals in which the effects of minor amounts of oxide and nitride can be determined and (2) for use as feed material in the production of low-oxygen-content yttrium metal, million, Solubility of Nitrogen and Oxygen in Lithium In order to determine whether cold trapping and low-temperature filtration would be practical methods for reducing the nitrogen and oxygen con- centration of lithium, several experiments were conducted to obtain solubility data. Preliminary data were obtained first on the rate of solution and the equilibrium solubility of lithium nitride UNCLASSIFIED ORNL—LR— DWG 30414 THERMQCOUPLE PROBE (IRON) = F ACE MUFFLE Mo SAMPLING URNAC BUCKET AND BAIL —— Mo CRUCIBLE Fig. 1.2.1. Apparatus for Measuring the Solubility of LigN in Lithium in a Purified Argon Atmosphere. in molten lithium at 752°F., The test was con- ducted, in a purified argon atmosphere, in the apparatus shown in Fig. 1.2,1. Specially prepared lithium nitride was fused in the bottom of the molybdenum crucible and allowed to solidify prior to the introduction of the lithium metal. The lithium nitride was fused prior to the test to reduce the pickup of particulate lithium nitride during sampling. The results of the test, as tabulated below, indicate that the solubility of nitrogen at 752°F (400°F above the melting point of lithium) is quite high: Time Nitrogen (min) (wt %) 0 0.11 30 0.99 60 1.50 180 2.10 240 2,86 300 3.14 In order to reduce the possibility of trapping impurity particles in the samples and yet obtain a large amount of sample for more complete analyses, a more refined solubility test rig was designed and built. This stainless steel apparatus is shown in Fig. 1.2.2. The test unit was loaded in a purified inert atmosphere with 60 g of lithium metal and sufficient lithium nitride or oxide to give a total oxygen or nitrogen content of 5 wt %. The system was sealed under 1 atm of pressure and placed in a furnace. At the end of the test run, the system was evacuated to a pressure of less than 5 p. The test unit was then tilted, as shown in Fig. 1.2.2, to submerge the filter head. Argon gas pressure was then applied to the system to force lithium into the evacuated tube, which was also at the test temperature. In order to avoid segregation of the impurities during solidification, the sampling tube was plunged into an oil—dry-ice bath. The stainless steel sampling tube was replaced with sampling ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 24830 VACUUM SYSTEM SHOWN TILTED APPROX. 30 deg TO HORIZONTAL BEFORE PRESSURIZING LITHIUM INTO SAMPLING TUBE AN » Iy \ E | > | | i 5 20- 4 STAINLESS STEEL FILTER 4] : & 4in 3 | | ‘ N N i i §ooil \\‘s j,// Mi,0 OR LigN [ ~—— SAMPLING TUBE Fig. 1.2.2. Stainless Steel Test Rig for Determining the Solubility of Oxides ond Nitrides in Lithium. one made of titanium for tests of oxygen solubility to facilitate activation analysis. The results of these tests are given in Table 1.2.1. These preliminary data indicate that low-temperature filtration and cold trapping are not promising methods of lithium purification if the nitride or oxide impurity content is to be reduced to below several hundred parts per million. Fiftration Early filtering experiments conducted by GE- ANPD indicated, in agreement with the conclusion above, that low-temperature filtration would not reduce the nitrogen and oxygen content.?2 Also, ]G.Leddicotte, Anal. Chem. Ann. Prog. Rep. Dec. 31, 1957, ORNL-2453, p 30. 2A. R. Crocker et al., Design, Operation, and Test of A Sodium-to-Lithium-to-Air Heat Transfer System, APEX- 178, p 112; APEX-327, p 117 (Dec. 1954). 10 the Nuclear Development Corporation of America (NDA) has reported unsuccessful attempts to purify lithium by filtration at temperatures only slightly above the melting point.® Filtration experiments conducted at ORNL have further substantiated these findings, as shown in Table 1.2.2. The slight increases in nitrogen and oxygen content upon filtering may be attributed either to inhomogeneities in the lithium or to contamination during or following sampling. It appears therefore that low-temperature filtration and cold trapping may be beneficial only in reducing the nitrogen and oxygen content of lithium that is very highly contaminated with nitrogen or oxygen. Yacuum Distillation Vacvuum distillation of lithium as a means of reducing the nitrogen and oxygen content has been studied at ORNL and at NDA. The NDA system has been used successfully to reduce the nitrogen content to below 10 ppm. In the NDA system, titanium sponge is added to the stifl prior to distillation to getter the nitrogen and to assist in the purification. No titanium was added to the still in the distillation performed at ORNL. The ORNL distillation system and allied equipment are shown in Figs. 1.2.3 and 1.2.4. Approximately 1100 g of lithium is charged to the still and 80% of each charge is distilled into the receiver. The remaining 20% is dis- charged through the drain line at the completion of each distillation. The distillation rates average 30 g/hr or 2.5 g/hr per square inch of liquid-vapor interface area. Six distillations have been performed to date, and the nitrogen and oxygen content of the distillate has been of the same magnitude as or higher than that of the lithium charged into the still in all distillations. The results of the six distillations are given in Table 1.2.3. The system operated in a satisfactory manner except as noted in the table. The operating conditions were identical with those used at NDA, insofar as possible, in order to ascertain the effectiveness of distillation when no titanium getter was added to the still. The results indicate that distillation without gettering is not a practical means of 3W. Arbiter and S. Lazerus, Purification of Lithium by Vacuum Distillation, NDA-39, p 7 (June 14, 1957). PERIOD ENDING MARCH 31, 1958 Table 1,2,1, Solubility of Nitrogen and Oxygen in Molten Lithium Filtering Aging Time Prior Test Solute* Temperature to Sampling Sample Solubility Remarks NO. (OC) (hr) No- (ppm) 1 Li3N 300 64 1 4,060 Samples slowly cooled; considerable 2 7,560 segregation occurred 3 19,600 2 LizN 250 24 1 1,000 Samples quenched to prevent 2 920 segregation 3 Li3N 250 72 1 1,080 Samples quenched 2 1,380 3 1,400 4 Li3N 300 24 1 1,760 Samples quenched 2 1,900 5 Li20 250 24 1 19,700 Samples quenched 2 22,700 *The test chamber was loaded with 60 g of |ithium metal lithium used contained 1076 ppm N, and 550 ppm 02. Table 1,2,2, Results of Tests of Low-Temperature Filtration as a Means of Purifying Lithium Impurities? (ppm) 02 N2 C As-received |ifhiumb Sample 1 334 560 15 2 446 645 Filtered tithium® Sample 1 595 680 11 2 518 760 18 2Analytical methods: (1) activation analysis for oxe ygen, (2) steam distillation followed by colorimetric analysis for nitrogen, and (3) spectrophotometric deter« mination of lithium carbide in metallic lithium as the acetylene—~silver perchlorate complex. Packed by vendor in stainless steel gastight cone tainers under helium. “Transferred through S-u stainless steel filter at 300°C following aging at 300°C. and 5 wt % nitrogen (as Li3N) or oxygen (as Li20). The purification. Lithium samples from all the purifi- cation tests are being exchanged between NDA and ORNL in order to check the reproducibility of the analytical results at the two installations. Gettering with Zirconium, Titanium, and Calcium The results of the filtration, distillation, and solubility tests described above indicated that gettering of the nitrogen and oxygen by active metals might be the best method of purification. The free energies of formation of various oxide and nitride compounds of interest were therefore examined (Table 1.2.4) in order to select the gettering metals to be tested. As may be seen, a considerable number of metals would probably getter nitrogen from lithium metal, but the oxide of lithium is extremely stable and only a few metals, among them calcium, yttrium, thorium, and beryllium, have oxides whose free-energy values indicate that they might be suitable as getters for reducing the oxygen content of lithium. Studies of the purification of lithium by gettering have thus far been limited to tests with zirconium, titanium, and calcium, 11 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 22264 R FULTON BELLOWS VAL\!"E] PRESSURE GAGE _AVACUUM LINE { Y4-in. TUBE) PRESSURE LINE __—6-in. FLANGE AND O-RING —g-in. TUBE -MICRO METALLIC FILTER 104 £, _—Yo-in. SCH 40 PIPE i_, 2 —THERMOCOUPLE GLAND DISCHARGE LINE TO STILL of : GAS FILL LINE ———— ADJUSTABLE LIQUID LEVEL PROBES L GAS RESERVOIR 110w TO AUTOMATIC SAFETY ___-SOLENOID VALVE (NORMALLY CLOSED) NOTE! VACUUM LINES ARE 1¥in. COPPER TUBING ; ALL OTHER MATERIAL TYPE 347 STAINLESS STEEL. VACUUM SYSTEM VALVES ARE 13g-in. BELLOWS-TYPE VEECO. FILTER Tt i ALPHATRON FILTER TANK TANK FiLL 19g-in. VEECO o J VALVE IBRASSy VACUUM GAGE SQ0-WATT i HASTINGS ELEMENT THERMOCOUPLE GAS- . . I"‘\ n.‘l o ' 570 | FURNACE — lasoec) ~ADJUSTABLE LIQUID — i £ 4-in, ISOTHERMAL |-0] | SODIUM JACKET — 7|8 3-in. BOILER POT—+H+ CLAM SHELL HEATER— | t-in. CONDENSER ,;" THERMOCOUPLE (675°C)- NE 4 (250 WATTS EACH) coLD TRAP = . LI i n DRAIN LINE- 3-in. RECEIVER P't:lT-’"gfi i i o A LEVEL PROBES (2) THERMOCOUPLE GLAND T GAS VAPOR BAFFLE -4 STRIP HEATERS ~THERMOCOUPLE (190°C) [} M.C.F. CENCO-HYPERVAC—25 - T AGE TANK DRAIN LINE TO STORAGE S ROUGHING PUMP DIFFUSION PUMP Fig. 1.2.3. Diagram of Lithium Distillation System. STILL — FURNACE UNCLASSIFIED 0 UNCLASSIFIED ¥Y=-23T78 CONDENSER RECEIVER Fig. 1.2.4. Lithium Distillation System With and Without Furnaces and Insulation. 12 PERIOD ENDING MARCH 31, 1958 Table 1,2,3, Effect of Vacuum Distillation at 650°C on the Nitrogen and Oxygen Content of Lithium Yacuum at cold trap with system cold: 2 x 10~ mm Hg Vacuum at cold trap with system ot temperature: 1 X 104 mm Hg Distillation No. Nitrogen* (ppm) Oxygen* {ppm) Remorks Charge to still (1100 g) 1170, 1250 226 As received and filtered ot 300°C through a 5-p filter 1 1400, 1300 370, 285 Av 1350 Av 327 2 1260, 1150, 885, 1010 700, 389, 376, 533 Av 1076 Av 500 3 3000, 1900 375 New still (4 in, pipe); condenser Av 2450 leaked near end of run 4 4200, 5300 1120, 1210 Some contamination of system occurred Av 4750 Av 1165 during repair of leak following dis- titlation No. 4 5 1600, 1600, 2000 203, 198 Av 1733 Av 200 6 1800, 2100, 2700, 3100 701, 758, 688, 630, Av 2425 Av 694 *Each value is for a separate analysis or is an average, as indicated, of the values given, Zirconlum Gettering Test. — Only one gettering experiment with zirconium has been conducted thus far. The appearance of the 0.022-in.-thick zirconium getter sheet before and after the test may be seen in Fig. 1.2.5, and the test results are summarized in Table 1,2.5. Metallographic sections of the zirconium specimen are shown in Fig. 1.2.6. The results of chemical analyses of the lithium and the zirconium indicate that zirconium is an efficient nitrogen getter under the conditions of this test. The absence of an increase in the oxygen content of the zirconium is in agreement with the free-energy data given in Table 1.2.4. The zirconium sheet weighed 33.913 g before the test and 33.933 g after the test; that is, there was a net weight gain of 0.65 mg/in.2. Titanium Gettering Tests. — Three lithium purifi- cation tests were run with titanium foil as the gettering and two tests were run with titanium sponge. It was realized that much more efficient gettering would be obtained with sponge, because of the larger surface area per unit of but foil was used in the preliminary agent, weight, experiments so that precise data on the gettering rates and the formation of oxide or nitride layers could be obtained by x-ray and metallographic examinations. The conditions of the first two tests with titanium foil were similar except for the test which was 1400°F in one test and 1500°F in the other. The test conditions and results are given in Tables 1.2.6 and 1.2.7. The results of both experiments show that titanium will getter nitrogen from lithium quite rapidly at both 1400 and 1500°F. The rate of nitrogen removal from the lithium was highest in the first 24-hr test period at 1500°F, as would be expected, since the diffusion rate of nitrogen into the titanium is greater at the higher temperature. The considerable decrease in the rate of removal during the second 24-hr test period emphasizes the need for a high ratio of titanium surface to lithium volume. As would be expected from the free-energy data, titanium is not effective in gettering oxygen from lithium. The considerable in the oxygen content of the [ithium temperature, increase is not understood; but, since the analytical 13 ANP PROJECT PROGRESS REPORT Table 1.2.4. Free Energies of Formation of Various Oxide and Nitride Compounds of Metals of Interest as Getters in the Purification of Lithium —AF® (kcal per gram atom of nonmetal) Metal Compound - At 25°C At 223°C At 727°C Lithiom Lizo(“) 138 130 123 - LioN(®) 37 Calcium ca0fc) 144 139 128 1 (5) Yttriom % ¥,0,(4) 143 yN(®) 64 Thorium % Tho, () 140 136 125 ] (b) Beryllium Be0(©) 139 134 123 ] (b) BeaN, 61 Uranium %u0,() 128 125 117 un(®) 75 . Barium Ba0(c) 126 120 107 i b) Y BagN, 37 Zircon ] 1 irconium /22r02 124 119 108 zeN(D 75 71 60 Titanium %y Ti0, 8 101 98 88 TiN(®) 73 Columbium I/Cb 0 (c) 85 5 275 CbN(®) 53 1 (c) Molybdenum 5Mo0, 60 Mo,N(6) 10 (@) E. Weber and L. F. Epstein, J. Met. and Ceramics, Issve No. 3, TID-67, p 85-110 (May 1949). (b)L. Brewer et al., Natl. Nuclear Energy Ser. Div IV 19B, 44 ff. (1950). ) (C)F. D. Rossini et al., Natl. Bur. Standards(U.S.) Circ. 500 (Feb. 1, 1952). (d)i. E. Campbell (ed.), High-Temperature Technology, Wiley, New York, 1956. ]95:6)0. Kubaschewski and E. L. Evans (eds.), Metallurgical Thermo Chemisiry, vol |, Academic Press, New York, (f/)B. Lustman and F. Kerze, Jr. (eds.), The Metallurgy of Zirconium, McGraw-Hill, New York, 1955. (S)F. D. Richardson and J. H. E. leffes, |. Iron Steel Inst. (London) 160, 261 (1948). 14 PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED Y-24728 0L\ 1] (o]0 Fig. 1.2.5. Zirconium Sheet (a) Before and (b) After Exposure to Lithium at 1400°F for 24 hr. Surface colors on the zirconium specimen after the test were bronze and purple. UNCLASSIFIED Y-25514 'rs:nx E ]E E ml:HE:s Table 1.2.5. Results of Lithium-Zirconium Gettering = T T : Test Ne. 1 & ) 2 Volume of lithium: 23.4 in.> B Weight of lithium: 180 g Surface area of zirconium sheet: 30.6 in.? Ratio of zirconium surface arec to lithium volume: 1.31in.%/in.3 Container: Armco iron ™ ! N i o i T -~ ..'I' ; . Test temperature: 1400°F . fl{-' . i .é.' = ' PR, (Hp e . . Impurity Content : E ZrN D : X Yo G.z'__ b LG 2 it (ppm) 5P aterial Analyze = ok # t\“‘ 2 02 BN OF N ( L Lithium Before test 1100 SURFACE BEFORE "|"| SURFACE AFTER EXPOSURE EXPOSURE After 24 hr 630 Zirconium Fig. 1.2.6. Metallographic Sections of Zirconium Spec- imens (@) Before ond (k) After Exposure to Lithium for Before test 70 1000 24 hr ot 1400°F. Note 0.0002-in. layer of Z¢rN on surface After 24 hr 510 1000 of exposed specimen. As polished. 750X, Reduced 28%. 15 ANP PROJECT PROGRESS REPORT Table 1.2.6, Results of Lithium=-Titanium Gettering Test No. 1 Test temperature: 1400°F Volume of lithium: 103.9 in.3 Weight of lithium: 800 g Surface area of titanium foil; 280 in.2 Ratio of titanium surface area to lithium volume: 2.69 in.z/in.a Container: type 347 stainless steel Weight of titanium foil: Before test: 128.910 g After test: 129.578 g Net gain: 0.668 g or 2,38 mg/in.2 Impurity Content Material Analyzed (ppm) Remarks N, O, Lithium Before test 3100 200 After 24 hr 1600 1230 Rate of N2 loss (0—24 hr), 62 ppm/hr After 48 hr 900 2280 Rate of N, loss (24—-48 hr), 28 ppm/hr Titanium foil (0.012 in. thick) Before test 73 1500 After 24 hr 1800 1400 X-ray examination of surface showed Ti, Ti{N), and TiN; metallographic examination showed 0.0002-in. layer of TiN After 48 hr 2000 1500 Xeray examination of surface showed Ti, Ti(N), and TiN; metal lographic examination showed 0.0004+in, layer of TiN procedure {activation analysis}) employed for determining the oxygen content is difficult and segregation of impurities in the test samples is a problem, it is possible that the increase in the oxygen content is an analysis error. Material balances indicate that the increased nitrogen content of the titanium accounts for only approxi- mately 40% of the observed weight increase of the foil. The surface color of the titanium was bronze following 48 hr of exposure to lithium at 1400°F and indigo following 48 hr of exposure at 1500°F. Polished and etched cross sections of the titanium specimens following the various tests are shown in Fig. 1.2.7. In test No. 3, which was run at 1500°F for a total of 192 hr, the stainless gettering container 16 was loaded with 280 in.Z of titanium foil that remained in the lithium for the entire test. This foil was wrapped on a stainless steel holder as in Fig. 1.2.8. Small titanium tabs with 6-in.?2 surface areas were inserted into the molten shown lithium during certain time periods of the test run as given in Table 1.2.8. The purpose of this was to determine the effectiveness of a fresh getter surface as the nitrogen content of the lithium decreased. The analytical data obtained from this test are presented in Tables 1.2.8 and 1.2.9. As may be seen, the nitrogen content of the lithium was lowered significantly during the experiment, and the rate of nitrogen pickup by fresh titanium did not appreciably as the nitrogen content of the lithium decrease PERIOD ENDING MARCH 31, 1958 Table 1.2.7. Results of Lithium=Titanium Gettering Test No. 2 Test temperature: 1500°F Volume of lithium: 94.8 in.® Weight of lithium: 730 g Surface area of titanium foil: 283 in.2 Ratio of titanium surface area to lithium volume: 2.98 in.zfin.a Container: type 347 stainless steel Weight of titanium foil: Before test: 129,287 g After test: 130.292 g Net gain: 1.005 g or 3.55 l'rlg,,i"in..2 Impurity Content Material Analyzed (ppm) Remarks Ny 0y Lithium Before test 2450 370 After 24 hr 740 910 Rate of N2 loss (0=24 hr), 71 ppm/hr After 48 hr 380 730 Rate of N, loss (24-48 hr), 15 ppm/hr Titanium foil (0.012 in. thick) Before test 73 1500 After 24 hr 1600 1000 Xeray examination of surface showed Ti(N) and TiN; metallographic examination showed 0,0004-in. layer of TiN After 48 hr 3200 1400 X-ray examination of surface showed Ti(N) and TiN; metallographic examination showed 0.0006~in, layer of TiN UNCLASSIFIED Y-25M3 BEFORE 24 hr AT 1400°F 48hr AT 1400°F 24 hr AT 1500°F 48 hr AT 1500°F Fig. 1.2.7. Surfaces of Titanium Gettering Specimens Before and After Exposure to Molten Lithium. Exposed surface at bottom. Note titanium nitride layer. Etchant: HF-HNO,-glyceria. Reduced 28%. 17 ANP PROJECT PROGRESS REPORT dropped from 550 to 100 ppm. At some nitrogen content in the lithium of less than 100 ppm very little titanium nitride formed on the fresh titanium surface (see 120 to 192 hr specimen), and, instead, massive quantities of nickel were picked up by the titanium from the lithium. Titanium tabs inserted at various periods are shown in Fig. UNCLASSIFIED Y-25444 . Fig. 1.2.8. Titanium Gettering Foil ond Stainless Steel Support Assembly for Foil. 1.2.9; the structure change resulting from nickel alloying may be seen. |t appears therefore that stainless steel is probably not a satisfactory container material for purification tests. The nickel, iron, and chromium content of the lithium samples was lower than expected at the 1500°F sampling temperature. |t has been reported that Table 1.2.8. Analyses of Lithium Samples from Lithium-Titanium Gettering Test No. 3 Test temperature: 1500°F Volume of lithium: 102.3 in.3 Weight of lithium: 788 g Surface area of titanium foil: 280 in.2 Ratio of titanium surface area to lithium volume: 2.73 in.z.a*’in.3 Container: type 347 stainless steel Impurity Content (ppm) Lithium Samples N, 0, Fe Ni Cr Before test 1400 After 24 hr 550 After 48 hr 350 1080 40 90 49 After 72 hr 200 2790 32 70 23 After 96 hr 100 1750 61 58 47 After 120 hr 125 694 29 65 13 After 192 hr 53 675 Table 1.2.9. Analyses of Titanium Samples from Lithium-Titanium Gettering Test No. 3 Impurity Content (wt %) Weight Gain of Titonium Samples Foil During N, 0, Fe Ni Cr Exposure (mg/in. Before test 0.044 0.14 0.17 0.017 0.010 0-24 hr 0.250 0.16 0.19 0.025 0.014 1.8 24-48 hr 0.140 0.17 0.18 0.027 0.012 0.9 48-72 hr 0.180 0.16 0.35 0.039 0.019 1.2 76-92 hr 0.140 0.14 0.16 0.033 0.014 1.3 96-192 hr* 0.200 0.17 0.21] 0.90 0.022 12.4 120-192 hr** 0.061 0.16 0.25 10.10 0.117 57.9 0-192 hr 0.3%90 0.15 0.26 0.23 0.0M 4.9 *Note high nickel content and large weight gain during exposure period. **Note low nitrogen content, high nickel content, high chromium content, and large weight gain. 18 PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED ¥- 25451 ANALYSIS OF SPECIMEN CROSS SECTION (wt %) N-0.25 0—-0.16 Ni—=0.02 Cr—0.01 Ti— BAL ‘ N—0.06 0- 0.16 Ni—40.10 Cr—0.42 Ti —BAL 4 b B o - e ,I:l’ ??;'—l*;\ 2 Woh sl 0.012 in. N-0.39 0—-0.15 Ni—-0.23 Cr—0.01 Ti—BAL Fig. 1.2.9. Titonium Specimens from Lithium-Titanium Gettering Test No. 3. (2) Tab exposed during 0- to 24-hr test peried. (b) Tab exposed during 120- to 192-hr test period. Note extensive phase changes as a result of high nickel pickup. (c) Specimen exposed during 0- to 192-hr test perioed. Etchant: 25% HF -25% HN03—5D% glyceria. Reduced 20%. 19 ANP PROJECT PROGRESS REPORT the solubility of nickel in lithium is 1580 ppm at 1200°F.* Analyses of the stainless steel container showed (Table 1.2.10) that stainless steel does not contribute significant amounts of nitrogen to the fithium. There was some mass transfer of the titanium to the stainless steel container wall. Table 1.2.10. Results of Analyses of Type 347 Stainiess Steel Container Used in Lithium-T itanium Gettering Test No. 3 Impurity Content {(wt %) Description of Sample 02 N2 Ti Before test 0.048 0.086 0.032 After test 0.050-in. cut from 0.045 0.074* 0.055 inside wall 0.050-in. cut from 0.029 0.038* ocutside wall *Oxide scale on outside removed before sample taken for analysis. Approximately 20 kg of badly contaminated lithium was recently received from Nuclear Development Corporation of America. This lithium had been exposed to air in 2-Ib lots for a consid- erable period of time. Of this material, 10.4 kg was pressurized through a 65-u stainless steel filter at 250°F into a stainless steel container. A 4-in.-dia high-temperature finger containing 1.7 kg of titanium sponge extended from the bottom to the top of the 12-in.-dia stainless steel container, The purification tank, the lithium, and the hot finger were maintained for 100 hr at T000°F and then for another 100 hr at 1400°F. The nitrogen content of the lithium after filtering and before gettering was 2250 ppm and after the gettering was 33 ppm. The oxygen content, which was 950 ppm after gettering, was not substantially altered by the purification treatment. Lithium-Titanium-Calcium Gettering Test. — As indicated above, the purification of lithium by 4. Q. Bogley and K. R. Montgomery, The Solubility of Nickel in Lithium, IGR-TN/C-250 (Sept. 30, 1955). 20 gettering the nitrogen and oxygen with titanium has only been partially successful. Titanium has been an excellent getter for nitrogen in all tests thus far, but it seems doubtful that it will be wuseful in reducing the oxygen content of lithium. In an effort to prepare low-oxygen-content lithium for use in yttrium metal production, gettering with both titanium metal sponge and triple-distilled calcium was tried. Calcium is one of the few metals whose oxide has a lower free energy than that of lithium oxide (see Table 1.2.4). There is considerable disagreement between the two available calcium-lithium phase diagrams, par- ticularly with regard to the solubility of calcium in lithium at temperature slightly above the melting point of lithium, At 400°F for instance, Zamotorin® gives the solubility of calcium in liquid lithium as approximately 3%, while Wolfson® gives a value of approximately 60%. The conditions and results of the first test with titanium ond calcium are given in Fig. 1.2.10. As may be seen, the purification procedure reduced the nitrogen content to a very low value but the oxygen content actually increased. The fairly low calcium content (1.95%) of the lithium product as compared with the amount (10%) placed in tank No. 1 indicates either that the solubility of calcium in liquid lithium is lower than the valué given by Wolfson or that the calcium alloyed with the titanium in tank No. 2 during the high- temperature gettering treatment, The latter pos- sibility is thought to be the more probable. STATIC LITHIUM CORROSION OF YTTRIUM E. E. Hoffman An yttrium metal specimen was exposed to static lithium for 100 hr at 1500°F in a columbium container tube. The test assembly and the con- siderable amount of yttrium metal which mass- transferred from the test specimen to the inner wall of the columbium tube are shown in Fig. 1.2.11. The yttrium cylinder decreased in diameter from 0.210 in. to as low as 0.207 in. in some areas, and it lost 0.10 g/in.2, or 2.4%, in weight. The extensive attack of the yttrium is illustrated M. I. Zamotorin, Metallurg 13(1), 96—99 (1938). 6M. R. Wolfson, Trans. Am. Soc. Metals 49, 794 (1957). LITHIUM CHARGE 840 g Oo,—= 730 ppm No — 2900 ppm PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED ORNL—LR—-DWG 31097 GETTERED LITHIUM 0, —1040 ppm N, —102 ppm —1.95% LITHIUM SAMPLING TUBE ——"fl-——|_—C° 95% TANK NO. TANK NO. 2 TANK NO. 3 Z-—:—— 84 g Ca \ N, 240 g Ti S SPONGE CHARGED WITH LITHIUM HELD AT 1200°F -1 hr PRESSURIZED INTO TANK NO. 2 AT 750°F CHARGED WiTH CALCIUM— LITHIUM ALLOY HELD AT 1400°F ~12 hr HELD AT 1000°F —80 hr STORAGE TANK PRESSURIZED INTO TANK NO. 3 AT 600°F Fig. 1.2.10. Test Conditions and Results of Lithium-Calcium-Titanium Gettering Test No. 1. in Fig. 1.2.12, which shows the exposed end of the yttrium cylinder following the test. The specimen, test, was heavily etched by the lithium, par- ticularly along the grain boundaries. [t appears therefore that the solubility of yttrium in lithium at the test temperature of 1500°F is appreciable. A metallographic section of the yttrium specimen is shown in Fig. 1.2.13. A phase, which has tentatively been identified as yttrium oxide, may be seen along the surface and in the grain boundaries. The bulk oxygen content of the yttrium specimen increased from 1900 ppm before the exposure to lithium to 3000 ppm following test. The possibility of using yttrium metal as a gettering agent to reduce the nitrogen and particularly the oxygen concentration of the lithium to be used in the subsequent production of low-oxygen-content yttrium metal is being considered. which was machine-ground prior to STATIC LITHIUM CORROSION OF ZIRCONIUM-BASE ALLOYS E. E. Hoffman Difficulty has been encountered ductile saddle the use of columbium weld-filler rod. in making columbium welds with Therefore two zirconium-base alloys were tested for use as welding-rod alloys in the fabrication of various columbium corrosion test components. Both an 85% Zr-15% Cb and an 80% Zr-15% Cb-5% Mo alloy were used, and the binary alloy was found to yield the most ductile, crack-resistant weld- ments. The details of these welding studies have been reported,” and the results of static corrosion tests are given below. In addition, tubing ’R. L. Heestand and E. E. Hoffman, Welding of Columbium. Part 1, ORNL-2423 (to be published). 21 ANP PROJECT PROGRESS REPORT ‘ COLUMBIUM TUBE LITHIUM LIQUID LEVEL 1 ‘.-—- YTTRIUM SPECIMEN Fig. 1.2.11. 1500°F. Reduced 18%. Fig. 1.2.12. Surface of the End of an Yttrium Cylinder Following Exposure to Lithium for 100 hr ot 1500°F. Reduced 34%. (Secret with caption) 22 ‘,'.s - & gl b -:;.0 “‘ -"i - ‘poe 0.25 in, el e T % ";. Moo v B Ne ?i:"‘;rfluu €230y alom, ";'a -, INNER WALL OF COLUMBIUM TUBE Yttrium-Lithium Corrosion Test Assembly ond Inner Wall of Columbium Container After 100 hr at several columbium thermal-convection loops have been fabricated and tested in which the saddle welds and butt welds were made with the 85% Zr-15% Cb alloy. During relatively short-time exposures to lithium in these loop tests (115 hr at 1500°F and 36 hr at 1600°F), weld joints made with the binary alloy were found to have excellent corrosion resistance to the lithium. Details of these tests are given in the following section. In the static tests, specimens of the B85% Zr—15% Cb alloy and the 80% Zr-15% Cb-5% Mo alloy were exposed to lithium for 100 hr at 1830°F. The ternary-alloy specimen showed no weight change as a result of the exposure, and the binary alloy showed a very slight weight gain (0.001 g/in.?). As may be seen in Fig. 1.2.14, neither specimen was ottacked. The 85% Zr-15% Cb alloy was used to seal the columbium corrosion test capsule, and the oppearance of the weld joint following the test is shown in Fig. 1.2.15. Some intergranular cracking occurred near the top of the weld joint as a result of the thermal treatment during the test. However, the test PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED Y-25489 || L 2 o INGI-llE‘B w " -4 - . . g “- .-“ . .-'. : F - . "N 3 ol ; ) i ; - 2 - 3 g W T e % " s 0 s .l-' : e J "4 . v T \.l.\- [_F b E * £ - o L <.\ e W s n A - g g - 3 : .WH *H‘_‘ ‘-L}“-“:"\ N .I - i ‘ ‘__ .. : : _.- i -'-‘-l ': .':: = wh_ . Las & I ] o o I s -: B f . 4; i “.‘. lfik.‘ -r ‘ ' T 5?0! Fig. 1.2.13. Metallographic Section of Yttrium Specimen Following Exposure to Lithium for 100 hr ot 1500°F. White phase along surface (top) and the grain boundaries has tentatively been identified as yttrium oxide. Etchant: HNDB-HC2H302-H20. 500X. (Secret with caption) UNCLASSIFIED ¥Y-25070 H&\E} é‘ ;- 003 - ~ w & 2 |fi "INCHES ~ I 1 1\ 2 2 - o o™ 80% Zr—15% Cb — 5% Mo I 250% 1 85% Zr —15% Cb ( Fig. 1.2.14. Metallogrophic Sections of Two Zirconium-Base Alloys Following Exposure to Lithium for 100 hr at 1830°F. Cathodic etch. 250X. Reduced 12%. (Secret with caption) 23 ANP PROJECT PROGRESS REPORT Zr-15% Cb ¥ % Ty A (o) 50 % - R Fig. 1.2.15. Section Through End of Columbium Tube Where 85% Zr—15% Cb Alloy Was Used os Filler Rod To Seal the Crimped Edges. Specimen was exposed to lithium for 100 hr at 1830°F. Enlarged view of cracked section in () may be seen in (k). As polished. Reduced 17%. (Secret with caption) capsule did not fail. Although most of the cracking occurred in the brittle grain-boundary phase, some transgranular cracking was also detected. The diamond pyramid hardness of the Zr-Cb weld joint was 236 as brazed, and the weld was ductile. The hardness was essentially the same following the test, 238 DPH, but the joint was found to be extremely brittle, as shown in Fig. 1.2.15. Although no attack was detected in the base material of the sintered columbium tube wall, heavy intergranular penetration by lithium was detected in the longitudinal tube weld and in the heat-affected zone of the weld. This attack was found to extend to a depth of approximately 0.025 in. THERMAL-CONVECTION LOOP TESTS OF LITHIUM IN COLUMBIUM E. E. Hoffman As was reported previously,® three of the four columbium thermal-convection loops that have 24 circulated lithium have shown heavy intergranular attack. These earlier tests and the tests to be discussed here were conducted in a stainless steel vacuum chamber. The procedure used, as illustrated in Fig. 1.2.16, permits the testing of unclad columbium loops. The attainable vacwum (less than 5 p with the loop at operating temperature) has been adequate to prevent oxi- dation or embrittlement of the columbium tube wall. The tests conducted to date have been of limited duration because of the high rate of penetration of the longitudinal tube wall weld by the lithium metal. The penetration is illustrated in Fig. 1.2.17(a), which is a picture of a section of tubing from the hot leg of loop No. 58, and in Fig. 1.2.17(b), which is a transverse section through the tube wall at the location of the leak. Both the loops (Nos. 58 and 59) to be described BE. E. Hoffman and L. R. Trotter, ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 173. \ | A1 CLAM-SHELL HEATERS | | | | | l | | i LOOSELY WRAPPED l HEATING ELEMENT | {TURNED OFF AFTER | FLOW HAS STARTED)'\f f ! 4 CRIMPED AND WELDED fi ’ B P { LA A7 e /‘: a5 \ i SEE SPECIMEN DETAIL L~ INSULATION PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED ORNL-I_LR-DWG 23499 SPECIMEN SUPPORT WIRE 0.035-in. THICK SPECIMEN DOOR — STAINLESS STEEL BOX— - Location of Loop in Vacuum Box. Fig. 1.2.16. Thermal-Convection Loop in Yacuum Chamber. here were constructed of I/2-in.-OD, 0.035-in.-wall sintered columbium tubing and each contained an insert of arc-cast columbium tubing in the hot-leg section, as shown in Fig. 1.2.18. The tests of both loops were terminated by lithium penetration of the longitudinal tube wall weld. The extent of the attack in the hot and cold of these loops is given in Table 1.2,11. legs Very little attack was detected in the base material of the tube wall away from the weld and the heat-affected zone of the weld. In general, the intergranular attack in the heat-affected zones adjacent to weld areas was as deep as in the the weld zone itself. The metallographic sections showing the attack of the arc-cast insert specimen and of the sintered tube wall are presented in Figs. 1.2.18 and 1,2.19. A hardness survey (25-g load) was made on the tube wall, as shown in Fig. 1.2.18, in order to determine whether there were any areas of embrittlement in the tube wall which might indicate pickup of elements such as nitrogen or oxygen either from the lithium or the atmosphere (vacuum) during test. No such embrittlement was detected, and, in fact, the arc-cast specimen was softer following test than before. The mechanism of the attack detected primarily in the weld zones and heat-affected areas of these loops is not known at present, but it is thought to be due to precipitation or segregation of impurities into the grain boundaries during the welding operation. This does not, however, explain the heavy attack of the arc-cast columbium tube inserts in both tests. The sheet specimens that were exposed in the hot legs in the position shown in Fig. 1.2.16 were examined for weight change. The specimen in loop 58 showed a weight loss of 7.5 mg/in.?, and the specimen in loop 59 gained 0.62 mg/in.2. 25 ANP PROJECT PROGRESS REPORT i UNCLASSIFIED Y-24729 'UNCLASSIFIED Y-24788 VACUUM Fig. 1.2.17. Section of Sintered Columbium Tubing from Hot Leg of Loop No. 58 (See Fig. 1.2.18 for Over-All View). (@) Columbium tube showing location of lithium penetration through longitudinal weld. (&) Metallographic section of tube wall weld zone showing penetration along grain boundery. Etchant: HF-H2504-HN03-H20. Reduced 6%. (Secret with caption) Nitrogen and oxygen analyses of various sections of the columbium from loop No. 58 are given in Table 1.2.12. For these analyses, weld metal was cut from the tube wall and analyzed separately from the base material. The nitrogen and oxygen content was slightly higher in all the specimens tested than in the as-received material. Static capsule tests are presently being conducted on specimens of various grades of columbium sheet that were inert-gas welded with varying degrees of protection during welding to study the effect of pickup of contaminants such as oxygen and nitrogen on corrosion resistance. STATIC LITHIUM CORROSION OF COLUMBIUM E. E. Hoffman A series of static tests was conducted in order to determine whether the type of environment (vacuum, 1 to 5 ) in which the thermal-convection 26 loop tests were conducted was in any way re- sponsible for the intergranular attack of the weld and heat-affected zones of the loop material. Four sintered columbium tubes of the same material (N,, 0.016%; O,, 0.027%) as that used to construct loops 58 and 59 were loaded with lithium metal. Tweo tubes were then tested in a stainless steel container as illustrated sche- matically in Fig. 1.2.20(¢), and the other two tubes were tested in the vacuum chamber as shown in Fig. 1.2.20(4). The tests were conducted at 1500°F for 100 hr. Argon gas (160 liters) con- taining 5 ppm O, and 58 ppm N, was circulated through the test assembly shown in Fig. 1.2.20(a). A dynamic vacuum of less than 5 p was maintained around the test assembly shown in Fig. 1.2.20(b). At the completion of the tests, various sections of the test capsules were examined metallo- graphically. The bath and vapor zones of each of the four columbium tubes were found to have Table 1.2.11. PERIOD ENDING MARCH 31, 1958 Results of Columbium-Lithium Thermal-Convection Loop Tests Nos. 58 and 59 Loop No. Location of Specimen Metallographic Data E ) 58 Hot leg Arc-cast insert Attack to a depth af 7 mils Sintered tube Weld zone Attack to a depth of 7 mils in most areas, complete penetration in several areas Base material No attack Cold-leg sintered tube Weld zone Attack to o depth of 5 mils Base material No attack 59 Hot leg Arc-cast insert Sintered tube Weld zone Base material Cold-leg sintered tube Weld zone Base material Attack to a depth of 16 mils Attack to a depth of 10 mils in most areas, complete penetration in several areas Attack to a depth of 2 to 3 mils Attack to a depth of 11 mils No attack Table 1.2.12. Results of Nitrogen and Oxygen Analyses of Columbium Sections from Thermal-Convection Loop No. 58 Specimen Nitrogen Content Oxygen Content (wt %) (wt %) Arc-cast celumbium tubing insert As received 0.009 0.045 After exposure to lithium in hot leg ot 1600°F 0.023 0,054 Sintered columbium tubing As received 0.016 0.027 After exposure to lithium in hot leg ot 1600°F WGId zone 0-025 0-04] - Base material 0.027 0.035 After exposure to lithium in cold leg ot 1300°F Weld zone 0.023 0.039 Base material 0.023 0.038 27 8¢ i 4 i /1. SINTERED _ 210 - /1" TUBING P b 7 i | ! 7 LITHIUM || FLOW [ 85% ze-15% Cb /AN 7 [ ALLOY —*T‘—“ ! ,’/f ! | . ARC CAST ] 77 TUBING \1 P ] 85% Zr-15% Cb ALLOY HOT LEG —= € o™ \ o™ | COLD LEG Fig. 1.2.18. Columbium-Lithivm Thermal-Convection Loop Mo. 58 Which Operoted for 36 hr with the Hot Leg ot 1600°F and the Cold Leg at 1300°F. e ORNL-LR-DWG 31069 FO.002 in- DIAMOND PYRAMID HARDMESS BEFORE TEST DPH-85 EMLARGED VIEW OF SINTERED TUBE WALL SINTERED COLUMEBIUM TUBE WALL w73 s i ) by ® A - Nt .i - BT T7 B84 o ‘g'-. & DIAMOND PYRAMID HARDNESS BEFORE TEST DPH-10T ENLARGED VIEW OF ARC-CAST TUBE WALL ARC-CAST COLUMBIUM TUBE WALL All photomicrographs show the specimens in the as-polished, unetched condition. Reduced 51.5%. L¥0d3d S5342038d LOIrodd dNV i ORNL- LR-DWG 3106A E - N e = | | 1 r _i | SINTERED __}| f AL TUBING | ' e 1 V _ i ~— dl I > by 7a [H V. LITHIUM B /f (W1 v FLOW e Il ll 85% Fr- "5":"9 Ch I. A L o L 3 \_1. ALLOY £ ._ 1 Mt /A MRC CAST { TUBING SINTERED COLUMBIUM £1_85% Zr-15% Cb_ : ALLOY 7 . L TUBE WALL HOT LEG —= ! V4 - | ' od | ! od L 1444 (: -+ f—————0.04 in.————=] ~— COLD LEG ) [ i ."- ; : g WELD JOINT FOLLOWING \\ EXPOSURE TO LITHIUM \\ F s e f'«. ' SECTION A-A THROUGH SADDLE WELD AREA SHOWN ABOVE ARC-CAST COLUMBIUM TUBE WaLL Fig. 1.2.19. Columbium-Lithium Thermal-Convection Loop No. 59 Which Operated for 115 hr with the Hot Leg ot 1500°F and the Caold Leg at 1300°F, All photomicrographs show the specimens in the as-polished, unetched condition. Reduced 56%. b} m 0 o o m 4 o z o = > o ™ T L :'Hi ° “n o ANP PROJECT PROGRESS REPORT — ARGON 41 3-in. —dio STAINLESS STEEL CAPSULE ORNL—LR—DWG 30445 - VACUUM | | 1! I | STAINLESS STEEL | ' VACUUM CHAMBER 1 I (2X4 X 4ft) ' L | vacuum I j/VACUUM — I | — '-——/ ! - | - I p=— | — | - nlintl ! | | I | | | | | | || 1 STAINLESS STEEL | /STArNLEss STEEL | | | -] | ' | FURNACE MUFFLE —a 77 ? FURNACE MUFFLE —=| 77 77 | I 1 | | | | | | T T : T T | H H ) H H ! | ] : | | | u U I U U | M M : M M : i /‘ | /l | | j | | / 1 — . | | | A L j ARGON COLUMBIUM TUBES (a) Fig. 1.2.20. the Test Capsule on Cotrasion. at least 25 mils of intergranular penetration, and, in several cases, there was complete intergranular penetration in the longitudinal weld zone and in the heat-affected zone of the tube wall. The base material of each of the four tube walls in regions away from the weld zone showed no attack, ¥ The nature of the attack in the weld and in the heat-affected zones of the tube wall is illustrated in Fig. 1.2.21. As may be seen, the tube wall was completely penetrated. A grain-boundary phase seems to be associated with the attack along the grain boundaries. As yet this phase has not been identified; it may be either columbium oxide or columbium nitride. The grain-boundary phase is very difficult to retain during normal metallographic polishing, 30 COLUMBIUM TUBES () Columbium-Lithium Corrosion Test Assembly for Studying the Effect of the Environment Around and therefore special polishing techniques were used on these specimens. METALLURGICAL STUDIES OF LEAD-LITHIUM ALLOYS D. H. Jansen Experimental work on a lead-lithium alloy con- taining 0.69 wt % lithium, which is a potential shielding material, was completed. Details of metallurgical studies conducted on the alloy have been reported.? An experimental lead-lithium shield measuring 36 x 36 x 4 in. was prepared for shielding studies at the ORNL Bulk Shielding Facility. %E. E. Hoffman et al., Lead-Lithium Shielding Alloy — Metallurgical Studies, ORNL-2404 {in press). PERIOD ENDING MARCH 31, 1958 e ", : o e - HEAT-AFFECTED ZONE |~ ¥ - -~ - “® Zh 3 r\_\*\fl% = \)5\0“ \ ‘\ N 0.035in. {¢) OUTSIDE OF TUBE {b) INSIDE OF TUBE {a) TUBE WALL AT WELD-HEAT AFFECTED ZONE INTERFACE UMCL ASSIFIED UNCLASSIFIED Y.25448 — : Y-25447 J=—0.003 in. ——w=] f=—0.003 in. = Fig. 1.2.21. Columbium Tube Wall Following Exposure to Static Lithium for 100 hr ot 1500°F. (a) Unetched, (&) and (c) etched with HF-HNOS-HESOJ-HZD. Reduced 41%. (Secret with caption) 31 ANP PROJECT PROGRESS REPORT 1.3. WELDING AND BRAZING P. Patriarca G. M. Slaughter R. L. Heestand E. A. Franco-Ferreira Metallurgy Division OXIDATION OF COAST METALS BRAZING ALLOYS 52 AND 53 G. M. Slaughter The ductility of o brozed joint is greatly dependent upon its microstructure, that is, the type and distribution of intermetallics, eutectics, and primary solid solutions. The redistribution or removal of the microconstituents should therefore result in a change in ductility of the brazing alloy. Gross depletion of intermetallics from a brazing alloy fillet should provide a significant improvement of its ductility. The formation of a nickel-rich solid solution layer on the fillet surfaces of joints brazed with Coast Metals brazing alloys 52 and 53 during oxidation has been reported.! Room- and elevated- temperature bend tests on Inconel samples over- laid with approximately 0.020 in. of brazing alloy have indicated that this depleted layer possesses considerable ductility and may actually provide a cushion to absorb imposed stresses. This condition is illustrated in Fig. 1.3.1, in which a crack in the braze matrix may be seen to terminate at the solid-solution interface. Pre- liminary hardness measurements have indicated that the hardness of the depleted zone is approxi- mately 140 VHN as compared with an interior hardness of 700 VHN. Thus it appears that the ductility of the oxidized joints should improve during service. This condition should be beneficial in most applications. A further study was conducted to determine the rate of depletion of microconstituents during high- temperature oxidation. Ten-gram cast buttons of the Ni-Si-B alloy (Coast Metals alloy 52) and Ni-Cr-Si-B alloy (Coast Metals alloy 53) were made and oxidized ot 1100, 1300, 1500, 1600, and 1700°F for 24-, 50-, 100-, 500-, and 1000-hr periods. The depth of the depleted layer was determined metallographically for each sample. 1P. Patriarca, A. E. Goldman, and G. M. Slaughter, Afillfs Quar. Prog. Rep. March 10, 1956, ORNL-2061, p " 32 The weight-change and depletion-depth data ob- tained for alloy 52 are presented in Table 1.3.1, and the data for alloy 53 are given in Table 1.3.2. Plots of the data are shown in Figs. 1.3.2 and 133 It may be seen that the extent of depletion is significantly greater in alloy 52 than in alloy 53. Also, it may be seen that the depletion in alloy 52 would still continue after 1000 hr, while it appears that the depletion has leveled off in the case of alloy 53. However, in all cases the depth of depletion increased with increasing temperature. Hardness studies and chemical and spectrographic UNCLASSIFIED PHOTO Y-25177 o m—| T O £ A 3 . At . 0.02 o ’ WP E;-;‘.-,_' * “f oy v i & - -. '.l '_;E — . -t - . -” . = - 0.03 ° s > . - Q- * Q . 2 Fige 1.3.1. Broze Matrix Showing Crack Which Terminates ot Solid-Solution Interface. As polished. 100X. analyses in the different zones will be conducted in an effort to more clearly understand the nature of the oxidation. UNCLASSIFIED ORNL-LR-DWG 30446 0.030 _0.025 ) 1700°F Q Q ] Q 0.015 DEPLETION DEPTH {in 1500°F 0.005 1300°F 100°F 1400°F 0 100 200 300 400 500 600 TIME (hr) 700 800 9S00 1000 Fige 1.3.2. Results of Oxidation Tests of Coast Metals Brazing Alloy 52. I PERIOD ENDING MARCH 31, 1958 DEVELOPMENT OF BRAZING ALLOYS FOR LITHIUM SERVICE G. M. Slaughter Refractory-metal-base brazing alloys are being investigated for a high-temperature reactor ap- plication in which they would be exposed to the lithium coolant. Nickel-base brazing alloys have been shown to have poor corrosion resistance to lithium, A literature survey was conducted to determine whether columbium and zirconium binary systems possessed eutectics or minimums. The results of the survey are summarized in Tables 1.3.3 and 1.3.4. A similar survey of the molybdenum, titanium, and tantalum systems will be conducted, and promising binary and ternary systems will be investigated in corrosion and flowability tests. It is expected that vacuum brazing will be required to obtain suitable flow of many of these alloys, and therefore a vacuum furnace system is being constructed. Table 1.3.1. Results of Oxidation Tests of Coast Metals Brazing Alloy 52 Weight Depletion Weight Depletion Temperature Time Temperature Time CF) (he) Change Dt.epth ©F) (hr) Change Depth (%) (in.) (%) (in.) 1100 24 +0.08 <0.001 1500 24 0.00 <0.001 50 ~0.12 <0.001 50 —0.02 0.001-0.002 100 +0.05 <0.001 100 —-0.04 0.002-0.003 500 +0.07 <0.001 500 +0.16 0.004-0.006 1000 +0.03 <0.001 1000 -0.22 0.007-0.009 1300 24 +0.09 <0.001 1600 24 +0.24 <0.001 50 +0.02 <0.001 ' 50 —-0.05 0.002-0.003 100 +0.11 <0.001 100 —~0.10 0.003-0.004 500 +0.18 0.001-0.002 500 —0.28 0.007-0.008 1000 +0.12 0.001-0.002 1000 —0.35 0.014-0.015 1400 24 —0.06 <0.001 1700 24 0.00 0.002-0.003 50 +0.10 <0.001 50 -0.15 0.003--0.005 100 +0.25 <0.001 100 —0.30 0.007-0.008 500 +0.02 0.001-0.002 500 ~1.00 0.018-0.020 1000 +0.14 0.002-0.003 1000 ~1.22 0.025--0.027 33 ANP PROJECT PROGRESS REPORT Table 1.3.2. Results of Oxidation Tests of Coast Metals Brazing Alloy 53 Temperature (°F) Time (hr) Weight Change (%) Depletion Depth (in.) 1100 24 +0.003 <0.001 50 +0.01 <0.001 100 +0.06 <0.001 500 +0.03 <0.001 1000 +0.18 <0.001 1300 24 +0.04 <0.001 50 +0.06 <0.001 100 +0.08 <0.001 500 +0.06 <0.001 1000 +0.10 <0.001 1400 24 +0.05 <0.001 50 +0.05 <0.001 100 +0.09 <0.001 500 +0.22 <0.001 1000 +0.22 0.000-0.002 (intermittent) 1500 24 +0.05 <0.001 50 +0.16 <0.001 100 +0.08 0.001 (intermittent) 1000 +0.006 0.003-0.004 (intermittent) 5000 +0.30 0.004-0.005 (intermittent) 1600 24 +0.15 <0.001 50 +0.08 0.001 (intermittent) 100 +0.17 <0.001 500 ~0.14 0.006~0.007 (intermittent) 1000 -0.12 0.005-0.007 (intermittent) 1700 24 ~0.007 0.002-0.003 (intermittent) 50 ~0.02 <0.001 100 +0.14 0.002-0.004 {intermittent) 1000 —-0.32 0.008-0.010 (intermittent) 5000 ~0.41 0.008-0.010 (intermittent) 34 PERIOD ENDING MARCH 31, 1958 The Zr-Cb and Zr-Cb-Mo alloys have been used Table 1.3.4. Zirconium Binary Systems of Potential as filler materials in the welding of columbium Interest as Brazing Alloys thermal-convection loops, and both alloys have shown excellent resistance to corrosion by lithium Eutectics or Minimums* (see Chap. 1.2, this report). Further studies of System Composition Melting Temperature these alloys will be conducted, and the Zr-Mo (wt %) CF) and Zr-Ti alloys, which should also be corrosion resistant, will receive early attention. Zr 3353 Zr-Al 11 Al, eutectic - 2460 UNCLASSIFIED 22 Al, eutectic 2700 ORNL -LR-DWG 30447 0020 ! 46 Al, eutectic 2700 £ oots - | R | Zr-B 2 B, eutectic 3200 E | . 38 B, eutectic 4208 S ' f700°F 75 B, eutecti 3380 g 0.010 |- 1 . /L____.—T —— 1 o y eutectic 3 F // 1600°F Zr-Be 5 Be, eutectic 1800 & 0005 |- 1 — o HB00°F - | oo e, eutectic 5 2 =" [1s00°F_ 93 Be, 2250 a . """ 1300°F o ok " | KES:)D'F HOOTF Ty, 1 Zr-Cb 25 Cb, minimum 3200 0 100 200 300 400 500 600 700 80O 900 {0CO Zr-Co 20 CO, eutectic 1750 TIME (re) 45 Co, eutectic 2750 65 Co, eutectic 2700 Fige 1.3.3. Results of Oxidation Tests of Coast Ze-Cr 18 Cr, eutectic 2340 Metals Brazing Alloy 53. 69 Cr, eutectic 2970 Zr-Cu 21 Cu, eutectic 1830 37 Cu, eutectic 1710 Table 1.3,3, Columbium Binary Systems of Potential 47 Cu, eutectic 1635 Interest as Brazing Alloys Zr-Fe 16 Fe, eutectic 1715 84 Fe, eutectic 2426 Eutectics or Minimums* ) Zr-Ge 7.7 Ge, eutectic 2795 System . Melting 99 Ge, eutectic 1712 Composition Temperature ‘ (wt %) CF) Zr-In 22 In, eutectic ~ 2550 Zr-Mn 22.5 Mn, eutectic 2076 Cb 4475 Zr-Mo 31 Mo, eutectic 2768 Cb-Co 79 Co, eutectic 2300 Zr-Ni 17 Ni, eutectic 1760 27 Ni, eutectic 1805 Cb-Fe 33 Fe, eutectic 2850 82 Fe, eutectic 2500 Zr-5b 22 5b, eutectic 2600 Cb'Ni 48 NI, eutectic 2]50 ZI"Si 2-4 Si, eU"eCfiC 2940 77 Ni, evtectic 2320 75 Si, eutectic 2470 Cb-Si 18 Si, eutectic 3450 Zr-Sn 23.5 Sn, eutectic 2894 58 Si, eutectic 3400 Zve-Th 75 Th, minimum 2426 95 Si, eutectic 2330 Zr-Ti 55 Ti, minimum 2900 Cb-Th 93 Th, eutectic 2600 Zr-V 30 V, eutectic 2246 Cb-v 65 V, eutectic 3290 Zr-W 18 W, eutectic 3000 Cb-Zr 75 Zr, minimum 3200 Zr-Zn 31 Zn, eutectic 1860 *J.P. Page, Alloys and Compounds of Niobium, ORNL *B, Lustman ond F. Kerze, Jr. (eds.), The Metallurgy CF-56-10-36 {Oct. 8, 1956). of Zirconium, 1st ed., McGraw-Hill, New York, 1955, 35 ANP PROJECT PROGRESS REPORT 1.4. MECHANICAL PROPERTIES D. A. Douglas Metallurgy Division STRAIN-CYCLING INVESTIGATIONS Investigations of the problem of thermal fatigue in reactor materials were continued. The capacity of Inconel to absorb strain reversals is being intensively studied in the temperature range from 1200 to 1650°F. as strain amplitude, temperature, cycle frequency, grain size, geometry, and environment are being considered. Three types of test have been em- ployed in these investigations: mechanical strain cycling, thermal strain cycling, and mechanical stress cycling (fatigue). The mechanical-strain- cycling tests are being conducted at ORNL. The other two test programs have been subcontracted to The University of Alabama (thermal strain cycling) and Battelle Memorial Institute (fatigue). Much of the information obtained thus far from mechanical-strain-cycling tests has been reported The effects of such variables in previous reports in this series.'=3 The con- clusions that have been drawn from the test results are (1) that the relation between total plastic strain absorbed per cycle (€.} and the number of cycles to failure (N} is of the form a — (n NeP—C, where a and C are constants; (2} that grain size is a significant factor, with fine-grained material being superior to coarse-grained material in its ability to survive a greater number of cycles; and (3) that temperature does not have a large effect on time to rupture in the range of 1300 to 1600°F at cycle frequencies of 0.5 per minute. Recent investigations have been concerned with the effect of the cycle time (frequency) on the relationship between € _and N. In these tests a load is applied which is of a magnitude that will cause the specimen to experience a specified plastic strain in one half of the specified cycle ‘C. R. Kennedy and D. A. Douglas, ANP Quar. Prog. Rep. June 30, 1957, ORNL-2340, p 190. 2c. R. Kennedy and D. A. Douglas, ANP Quar. Prog. Rep. Sept. 30, 1957, ORNL-2387, p 182. 3D. A. Douglas et al., ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 153, 36 R. W. Swindeman time, and then the load is reversed to produce - the same plastic strain in the opposite direction in the other half of the cycle time. Comparisons of data obtained at different frequencies and temperatures are presented in Figs. 1.4.1 through 1.4.4, As may be seen in Fig. 1.4.1, the results for specimens tested in argon at 1300°F with 2- ond 30-min cycles give €, vs N curves that fall close together. When the plastic strain per UNCLASSIFIED DRNL -LLR-DWG 30448 100 o Q ¢ 2-min CYCLES o 30-min CYCLES N o S i8] €, PLASTIC STRAIN PER CYCLE (%) Qo Mmoo~ N o o o 01 02 05 ¢ 2 5 1o 20 50 100 200 500 1000 N, CYCLES TO RUPTURE Fig. 1.4.1. Effect of Strain Cycling on Fine-Grained Inconel Tubes Tested in Argon at 1300°F. UNCLASSFIED ORNL-LR-DWG 30419 100 >0 2 min CYCLES {0-min CYCLES 20 20-min CYCLES 30-min CYCLES 05 ¢,, PLASTIC STRAIN PER CYCLE (%) 02 Oa 2 5 10 20 50 100 200 500 1000 10,000 N, CYCLES TO RUPTURE Figl ]-402. Inconel Tubes Tested in Argon at 1500°F. Effect of Strain Cycling on Fine-Grained UNCLASSIFIED ORNL-LR-DWG 30420 8 S o ® 2-min CYCLES 4 {0~ min CYCLES 4 20-min CYCLES © 30-min CYCLES n Qo o [6)] PLASTIC STRAIN PER CYCLE (%) O o LN v, o o o 1 2 5 10 20 50 100 200 500 1000 N, CYCLES TO RUPTURE 10,000 Fige 1:4.3. Effect of Strain Cycling on Coarse«Grained Inconel Tubes Tested in Argon at 1500° F. UNCLASSIFIED ORNL -LR-DWG 30421 100 50 —- n O O w\ - n €p, PLASTIC STRAIN PER CYCLE (%) o o 0.2 b2 5 40 20 50 N, CYCLES TO RUPTURE 100 200 500 1000 Fig. 1.4.4. Effect of Strain Cycling on Fine-Grained Inconel Tubes Tested at 1500, 1650, and 1900°F in Argon at 30 Minutes per Cycle. cycle is above 1%, the 30-min cycle results in a slight increase in the number of cycles before failure. The effect of long cycle periods at low strains (below 1%) remains to be clarified. In similar tests of coarse- and fine-grained specimens at 1500°F, cycling periods of 2, 10, 20, and 30 min were used. The results of these tests, as presented in Figs. 1.4.2 and 1.4.3, show, in all cases, that when the strain per cycle is PERIOD ENDING MARCH 31, 1958 above 5% the cycle frequency has no significant effect on the number of cycles to rupture. At low valves of €, however, the cycle time becomes increasingly important, At the same plastic strain per cycle, the number of cycles to rupture de- creases with increases in the cycle period. From a standpoint of actual time to rupture, however, long cycle periods result in increased life. Since the 30-min cycle period is 15 times greater than the 2-min cycle period, a considerable reduction in the number of cycles to failure would have to occur before the time to rupture would be shorter for the 30-min cycle than for the 2-min cycle. A comparison of results obtained at 1300, 1500, and 1650°F with a 30-min cycle is given in Fig. 1.4.4. The data indicate that the ‘'break point,”’ that is, the point where a frequency effect is first noticeable, moves to higher values of ¢ as the test temperature is increased. These data contrast with data for 2-min cycles, which showed little effect of temperature. The existence of a break in the € vs N curves is somewhat ques- tionable, since it is possible that a smooth curve which did not satisfy the empirical relation given by Eq. 1 could be drawn through the data. Actually, in the type of test system employed in these frequency investigations, the value of €, is not constant throughout the test; a small increase in €_occurs as the test progresses. In general, the change in € occurs in the first few cycles and thereafter increases only slightly. The values of € which are given in the data correspond to measurements taken in the middle or near the end of the test, and therefore the change in €, that occurs in the early part of the test is not reflected to a |arge extent in the curves presented in Figs. 1.4.1 through 1.4.4. Since the cyclic strains of greatest interest from a design standpoint are well below 5%, it appears that cycle frequency must be considered in evaluating the probable lifetime of a material that will be subjected to cyclic strains. Further work is needed therefore to establish the effects of strain cycling over a larger frequency spectrum and a larger strain amplitude range. Since the cyclic strains induced during the operation of a reactor are primarily the result of thermal rather than mechanical strain cycling, the mechanical-strain-cycling data can be useful in design only if correlation can be obtained with results from comparable thermal-cycling tests. Tests in which the strain is induced thermally 37 ANP PROJECT PROGRESS REPORT have been conducted at mean temperatures of 1300 and 1500°F with a cycle period of 38 sec. Data obtained at 1300°F in the two types of tests show excellent agreement when analyzed on the basis of the equation relating plastic strain to the number of cycles to failure.? Data obtained in similar tests at 1500°F are compared in Fig. 1.4.5. UNCLASSIFIED ORNL-LR-DWG 30422 o . MECHANICAVL- STRAIN - CYCLING DATA o THERMAL - STRAIN - CYCLING DATA 1 ey, PLASTIC STRAIN PER CYCLE (%) 10 20 50 100 200 500 1000 10,000 100,000 N, CYCLES TO RUPTURE Figs 1.4.5. Comparison of Effects of Thermal ond Mechanical Strain Cycling on Inconel Tested at a Mean Temperature of 1500° F. As may be seen, the correlation of the data from the two tests is poor, especially at values of ¢ greater than 0.1%. The lack of agreement may be attributed, in part, to difficulties in determining the plastic strain in fthe thermal-strain-cycling test. The specimens are resistance heated in the thermal-cycling tests, and high thermal gra- dients are set up that produce variations in the elastic and plastic properties. Therefore an ‘‘effective gage length’' must be established which may vary as the test progresses. Local hot spots produced by imperfections, cracks, or changes in cross-sectional area also produce localized strain. These effects are most pro- nounced for the high-temperature cycles, which correspond to the high-strain cycles. More refined methods of determining the plastic strain and of minimizing the temperature gradients are being considered. Some data have also been obtained concerning the effect of the cycle frequency on the life of material subjected to thermal strain cycling. Two tests have been conducted at 1300°F for which 30-min cycles were used. One specimen was cycled at 1.3% €pr and it survived 100 cycles 38 as compared with 600 cycles in a test for which the cycle period was 38 sec. The second specimen, which was cycled at 0.93% € , lasted 375 cycles as compared with 800 cycfes when tested on a 38-sec cycle. In some instances, service conditions may exist within a reactor in which the cyclic frequencies of the strains, whether mechanically or thermally induced, will be high, that is, 1 to 10 eps. High- temperature fatigue may occur under such con- ditions that is usually considered to be the result of stress cycling rather than strain cycling. Fatigue studies are therefore being conducted on Inconel specimens at 1200, 1400, and 1600°F. Results from tests at 1600°F in which frequencies of 1 and 10 cps were used are shown in Fig. 1.4.6. Under these conditions, the effect of cycle frequency is evident. Specimens cycled at 10 cps endure 10 times as many cycles at the same stress level as those cycled at 1 cps. An attempt is being made to dynamically measure the strain per cycle at these frequencies. If this can be done, the results of fatigue investigations can be correlated with the strain-cycling data. UNCLASSIFIED ORNL-LR-DWG 30423 16,000 = | ] ® | ‘ L . ¢ ! | : n 14,000 L N | e . & \c.;tcps ! \ -24 -6000,/6000 6000/0 0/6000 6000,/6000 STRESS RATIO, o, /o, Fige 1.4.15. Total Strain at Rupture of Inconel Tubes Creep Tested ot 1500°F Under Combined Axial and Tangential Stresses, UNCLASSIFIED ORNL-LR-DWG 30430 5 6000/0 L 6000 /2000 [ & 6000/4000 6000,/6000 4000/6000 o0 / / /e. STRESS RATIO, o, /o, 2000/6000 / 0/6000 - 20006000 / . -4000 /6000 / - 6000/6000 . 0 100 200 300 400 TIME TO RUPTURE ({(hr) Fig. 1.4.14. Time to Rupture as a Function of the Stress Ratioc in Inconel Tubes Bioxially Stressed and Creep Tested in Argon at 1500° F. 42 direction, which corresponds to the decreasing stress ratio, were to induce a compressive com- ponent to the creep in the axial direction, the observed creep rate would decrease as O’z/O’t decreased from = to 1. Such behavior would be predicted under the assumption that the sum of the strains and of the strain rates in the three principal directions must at all times be zero.® No influence of tangential creep is observed. Moreover, as the axial stress decreases and the stress ratio drops from 1.0, the axial creep rates remain in agreement with pure tensile data. No effect of the tangential strain is noted until the stress ratio approaches 0.5. Here, the sudden disappearance of axial creep can be partly at- tributed to foreshortening of the tube caused by ““bowing out” in the center of the gage length. Below the ratio 0.5, the axial creep rate in com- pression is higher than that which would be expected by assuming dependence on axial stress alone. Two additional factors influence this creep 8. Gensamer, Strength of Metals Under Combined Stresses, American Society for Metals, Cleveland, 1941. STRESS ({psi) 102 1078 2 5 1073 2 5 Tl 2 PERIOD ENDING MARCH 31, 1958 rate, however. One of these is the bowing-out effect previously mentioned. The other is an eccentric-load effect from the compressive axial load acting on the bowed-out tube. It is difficult to predict how creep in one direction may affect the deformation rate in the other., For one thing, deformation is accompanied by changes in the stress values and stress distribution, which, in turn, result in changes in the creep rates in the principal directions. Evidence of stress redis- tribution has been found during tests. Near the stress ratio of 0.5, specimens have been observed to start creeping in compression and then move into tension. The opposite effect has also been noted. On a microscopic scale, there are many factors which have not been considered in the theories of flow. For example, the generation of cracks, voids, and lattice defects by the maximum stress may influence the creep rate produced by a lesser stress in another direction. The stress de- pendence of the creep mechanism could also be considered. Preferred orientation, which is present to a degree in tube material, causes an UNCLASSIFIED ORNL—LR—DWG 30432 ay fop = © (UNIAXIAL) o‘z/o'f =2 z/% =1 05< az/crf(" c'z/crr: -0.5 —2 CREEP RATE (hr™ %) Fig- 104-]60 Tangential Stress Ratios. Axial Creep Rate of Inconel Tubes at 1500°F When Bioxially Stressed at Various Axial-to- 43 ANP PROJECT PROGRESS REPORT dependence of elastic ond plastic properties. Factors that influence the fracture point of the material are even more numerous. At the present time no clear fracture criterion has been established. orientation The maximum shear stress, 44 the octahedral shear stress, and the maximum normal stress theories of yielding have been rejected because they fail to explain the type of stress-ratio dependence which has been observed for the time to rupture. PERIOD ENDING MARCH 31, 1958 1.5. CERAMICS L. M. Doney R. L. Hamner R. A. Potter Metallurgy Division In cooperation with GE-ANPD, fabrication studies are being made of dense beryllium oxide for use as a neutron reflector material. The possibility of adding boron (0.5 to 1 wt %) as a suppressor for secondary gamma rays is also being inves- tigated. The two phases of the studies involve (1) a survey of boron compounds and their effects on the thermal stability and oxidation resistance of BeO ond (2) densification of extruded and cold-pressed BeO to which the boron compounds have been added. BORON COMPOUNDS The boron compounds of initial interest were the borides of the transition metals of the fourth, fifth, and sixth periodic groups, whose physical properties, such as thermal stability and oxidation resistance, made them particularly attractive for the application described. For preliminary studies it was necessary to synthesize the compounds in the laboratory. Diborides of Ti, Zr, Hf, Ta, V, and Cr were synthesized by reacting small batches (~ 100 g) of the metal oxides, or, in the case of zirconium, the metal silicate, with graphite and boron carbide in graphite crucibles heated by high-frequency induction to temperatures in excess of 2000°C. Because of reaction-product gases, it was difficult to determine the exact temperatures of the reactions by the available optical methods; therefore, visual observation of ignited out-gases and time-power control of the induction unit aided in determining reaction progress. The reaction between zircon sand, graphite, and boron carbide was more complete than the other reactions, and the product was cleaner. During the reaction, ZrB2 was produced as a mass of granules slightly sintered together which were similar in size and shape to the original sand particles. X-ray diffraction patterns were made of all the compounds, and in each case the material was found to be predominantly boride. BORON ADDITIONS TO BeO Specimens of boron-containing BeO were pre- pared for preliminary evaluation by mixing the boron compounds with Brush Beryllium Company “Luckey S.P."" grade BeO in sufficient amount to give 1 wt % boron and then hot pressing the mixtures at 1900°C at a pressure of 2000 psi. Specimens containing ZrB,, TaBz, TiBz, CrBz, HfBz, CeB,, and BN were prepared. Comparable densities (94 to 96% of theoretical) were obtained for all mixtures except CeBsoBeO, which was discarded because of its generally poor appearance. Microscopic examinations indi- cated that the additives were held in the BeO and uniformly distributed, with the exception of CrB,, the lowest melting compound of the series, which appeared to have migrated toward the periphery of the specimen. The fabricated specimens were oxidation tested at 1000°C and at 1300°C. At 1000°C the ZrB.,,- BeO and HfB,-BeO specimens showed no weigint change after 34 hr; the TaB,, TiB,, and CrB, combinations showed small weight changes (<0.2%) and attendant oxidation products. The exposed BN of the BN-BeO specimen had almost completely disappeared, and there were easily visible pits and craters. The disappearance of the BN is believed to be due to oxidation and subsequent volatilization. The ZrB, and HfB,, mixtures exhibited oxidation resistance at 1300°C that was much superior to the oxidation resistance of other combinations. The ZrB, ond HfB, specimens showed weight changes of only 0.15% after 300 hr. X-ray patterns of the ZrB,-BeO mixture before and after oxidation were identical. In view of its stability in air and its avail- ability, only ZrB, will be used in future work on this phase of the program. Specimens of the BeO-ZrB, mixture are being fabricated for determi- nations of modulus of rupture, compressive creep, and thermal conductivity before tests are made in the ETR. DENSIFICATION OF BeO Additives are being sought that will densify extruded and cold-pressed BeO bodies and possibly, but not necessarily, fulfill the boron 45 ANP PROJECT PROGRESS REPORT content requirements. Preliminary data obtained by the Brush Beryllium Company indicated that the following two compositions should be investi- gated: 94.0% BeO-5.0% MgO-1.0% B,C and 98.0% BeO-1.0% Fe,0,-1.0% B,C. An extruded body was produced from the first composition that, when fired to 1500°C in an oxidizing at- mosphere, had a density of approximately 2.86 g/cm?. No determination was made of the boron content. In order to determine effects of the individual components on BeO, mixtures were made con- taining the following parts by weight: 94.0 parts BeO plus 5.0 parts MgO; 94.0 parts BeO plus 1.0 part B,C; 98.0 parts BeO plus 1.0 part Fe,O,; 98.0 parts BeO plus 1.0 part B,C. Data obtained here have indicated that the additive ZrB, may possibly fulfill the density requirement, as well as the boron requirement, and, therefore, the 46 composition 94.8% Be0-5.2% ZrB, is also being investigated. The various compositions, including one of 100% BeO for comparison, were mixed in a jar on rolls for 18 hr and then pressed at 20,000 psi in steel dies, broken up, granulated by passing through a 30-mesh screen, and remixed for 2 hr. Right cylinder shapes approximately ¥ x ¥ in. were made by pressing the materials at 15,000 psi. Pieces from each of the compositions are being fired to 1100, 1300, and 1500°C in oxidizing atmospheres. These pieces will be compared with others fired in inert atmospheres and their dif- ferences in density and boron retention properties will be noted. In the future, these compositions, as well as additional ones, will be made into extruded shapes and compared with their cold- pressed counterparts. PERIOD ENDING MARCH 31, 1958 1.6. NONDESTRUCTIVE TESTING J. W. Allen R. W. McClung R. A. Nance Metallurgy Division EDDY-CURRENT THICKNESS MEASUREMENTS J. W. Allen R. A. Nance The study of eddy-current techniques for meas- uring cladding thicknesses on flat fuel plates was continued, with special emphasis on the ability of the probe coil to locate areas of minimum cladding thickness. graphic correlation studies,! sections have been taken through a point beneath the center of the probe, and the results have been erratic because the area of thinnest cladding does not necessarily occur at this point. In order to gain a better understanding of this problem, a study was made of the cladding thickness variations over the In past metallo- entire ]/2-in.-dia area underneath the probe coil. This was accomplished by successively removing thin layers of the cladding and plotting contours of the cladding thickness wvariations. Contour plots of two such areas which were examined are presented in Figs. 1.6.1 and 1.6.2. The several sections taken through the area beneath UNCLASSIFIED ORNL-LR-DWG 27533 PLAN VIEW CLADDING THICKNESS LEGEND CORE: 48 wt% U 3 wt% Si 49 wt% Al CLADDING: 100 Al 0.0065—-0.008n. 0.008—0.010in. LOCATION OF PROBE COIL 0.040—0.0115 in. (COIL DIAMETER %g in.} Ot W CLADDING THICKNESS MEASURED BY EDDY -CURRENTS = 0.014 in. 0.0115—-0.014 in. >0.014in, INCHES i i 3 cogs | colLoa “ ‘2 /4 : gy s CLADDING 0.010 [ F - A T e CORE :> 0_0’4 Y. //Il/ll/”l’l’/’/ll”’;”’. = SECTION A—A & 2 0006 [ LGOI DA ST "L~ CLADDING I o010 [= ® — CORE Zz 0044 = a Ld S 0006 o E CLADDING 0.010 ————— — BTt e s CORE A SLT LTS ILS S ST IR LSS ITS ST S ITS S ST IS Q.04 SECTIONC-C Fig. 1.6.1. Contour Plot of Clodding Thickness Yariations on Flat-Plate Fuel Element. the center of the probe indicate the varied meas- urements which can be obtained by changing the orientation of the section. Each section indicates a different minimum cladding thickness, none of which occurred beneath the center of the probe. The study of the possibility of measuring the thickness of a hot-dipped aluminum coating on steel was also continued. This metal system is not of any particular project importance in itself, but it is typical of metal systems in which a magnetic material is clad with a nonmagnetic material. The specimens that were examined had aluminum coatings of 0.0016-in. nominal thick- ness, including an intermetallic layer of approxi- mately 0.0003-in. thickness, as shown in Fig. 1.6.3. The intermetallic layer adversely affects the measurement of coating thicknesses of less than 0.0005 in., but the accuracy of the measure- ment is +0,0002 in. for the thickness range of 0.0005 in. to 0.0016 in. Measurements of aluminum coating thicknesses on steel in the absence of ]R. B. Oliver, R. A. Nance, and J. W. Allen, ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 192. UNCLASSIFIED ORNL-LR-DWG 27534 PLAN VIEW CLADDING THICKNESS LEGEND CORE: 49wt % U 3wt % Si 49 wt % Al CLADDING: 1100 Al c 0.0135—0.015in. ¢ LOCATION OF PROBE COIL 0.015 — 0,047 in. " {COIL DIAMETER %g in.) CLADDING THICKNESS >0.047 in. MEASURED BY EDDY-CURRENTS = 0.044 in. INCHES 0 Ya Ya Ya COIL DIA ———=q 0.013 T CLADDING - T Y £ o.oie B e CORE w SECTION A—A @ W z jw———COIL DIA——= s 0.043 RS ———_==- CLADDING I _.Il-l'l;};;l’l’;;;}lllltl’lfl: CORE — O X 04 e O e e B L L Ll L LD o SECTION B-B =2 2 o —— COIL DIA——= < 0.013 HT; T CLADDING © o e T 0.01g BRI CORE SECTIONC—C Fig. 1.6.2. Contour Plot of Cladding Thickness Variations on Flat-Plate Fuel Element. 47 ANP PROJECT PROGRESS REPORT 1 1‘&!0 : 3 S D W J L -.‘v.__:‘ ‘ Ah .1;\;\ 3 " b ‘3 i B e P ST UNCL ASSIFIED Y-24583 Fig. 1.6.3. Hot-Dipped Aluminized Coating on Low Carbon Steel. Etchant: 2% nital. an intermetallic layer can be made with an accuracy of 0.0001 in. or greater. While investigating the problem of measuring the thickness of aluminum coatings on steel, a new technique was developed which shows con- siderable promise of eliminating the inaccuracies introduced by lift-off or poor coupling between the probe coil and the part. Previously, thickness measurements have been made by measuring the amplitude variations of the impedance of the probe coil as it scans the part. These amplitude variations are determined with respect to the impedance of the probe when placed on a thick section of one of the component metals of the system being measured. This method requires that the surface of the part be flat and that the probe be in intimate contact with the surface or have a constant amount of lift-off. Even a small wvariation in lift-off (0.001 in.) introduces changes in both the amplitude and the phase angle of the coil's impedance, which greatly 48 affect the thickness measurement. |t was found that in measuring the thickness of aluminum on steel it is possible to eliminate the effects of lift-off by referring all impedance measurements to the impedance of the coil in air and then utilizing the phase change of the coil's impedance as a measurement of the thickness. With this technique, lift-off greatly affects the amplitude of the coil’s impedance but causes almost no change in its phase angle, as illustrated in the impedance diagram of Fig. 1.6.4. The phase angle, 8, of a vector from the air point to the thickness curve is used as a measure of the aluminum coating thickness. The introduction of lift-off causes the vector length to shorten, but has no appreciable effect on the angle 6. By using this technique the coating thickness measurements can be made to be largely independent of the amount of lift-off. A plot of the phase-angle variation as a function of aluminum coating thickness on steel is shown in Fig. 1.6.5. As in all eddy-current thickness UNCLASSIFIED ORNL—LR—DWG 26994 o Fe r— z 0.3 Ll = o a = C “ 7/ " > A 5 < 7 o ; P-4 @ : _- —— o )'/ 9‘ 5 Al THICKNESS ON Fe (in. x 1073) RESISTIVE COMPONENT Fig. 1.6.4. Normalized Probe Coil Impedance Vari- ations for Varying Thicknesses of Aluminum on Leow Carbon Steel at 20 kc Referenced to Impedance of Probe Coil in Air. measurements, the most sensitive measurements are those on the thinnest sections. At the present time the phase-angle measurement technique can be used to advantage only if the base material is magnetic. A study is being made to determine whether the technique can be successfully applied to systems having nonmagnetic bases. This method is particularly well suited to automation because of its independence of lift-off variations. LAMB ULTRASONIC-WAVE STUDIES R. W. McClung The detection of laminations in thin (]4 in. or less) metal sections and the nondestructive evaluation of bond quality in clad structures are among the more difficult inspection problems. Radiographic, eddy-current, and dye penetrant methods are not applicable to these problems because of unfavorable orientation and non- accessibility of laminar discontinuities. Con- ventional pulse-echo ultrasonic techniques do not have the requiredresolving abilities. Resonance ultrasonic methods can, in some instances, provide a partial solution to the problem; however, PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED ORNL—LR—DWG 26997 140 '\/fi} .—-—""——-—-—_—-—. % 100 [ - —pf— = }— w T 80 _— f— — — -+ — wn L _ g | Z 60 | — —;— - U_I ! @ | <{ | & 40 — - - ’—— - 20 _ o _ — J,_ ] | o 1 O 5 10 15 20 ——=o THICKNESS OF ALUMINUM ON LOW CARBON STEEL (in. x 107 3) Fig. 1.6.5. Eddy Current Thickness Gage Calibration Curve, 20 ke, the sensitivity of the resonance method is limited, and, being a contact technique, it is subject to the inherent difficulties of contact testing: relatively slow inspection speed and problems of contact pressure and application of couplings. A possible solution to these problems involves an ultrasonic phenomenon known as Lamb waves, which are basically a characteristic elastic vibration established in a thin metal plate. The vibration has a frequency in the ultrasonic and the frequency, plate thickness, and the mode of the vibration are interrelated. This phenomenon was postulated by Horace Lamb in 1916 on the basis of a mathematical study. A program of study has been initiated to gain fomiliarity with this generally unknown technique. The Lamb waves are being generated in thin metal sections by directing a beam of ultrasonic energy at particular incident angles to the sheet At the present time, commercial ultra- sonic instrumentation is being used to generate range, surface. and receive these vibrations. As more familiarity with the system is gained, it is planned to build specific which will optimize the test requirements. Early results in this study have been quite promising. instrumentation 49 ANP PROJECT PROGRESS REPORT ULTRASONIC ATTENUATION AND VELOCITY MEASUREMENTS J. W. Allen R. W. McClung The evaluation of any ultrasonic inspection result is difficult because of the attenuation of the ultrasonic energy in the part being inspected. This attenuation is almost entirely a function of the test metal structure and the frequency of the ultrasonic waves. The problem of ultrasonic attenuation is particularly troublesome in the inspection of weldments. In many instances, the attenuation is so severe that gross weld defects are not detectable. In order to overcome these problems of ultrasonic attenuation, much more information and knowledge must be obtained about the phenomenon. An ultrasonic attenuation com- parator has been procured for studying the atten- vation problem in more detail, and familiarization studies with this new equipment have begun. It is planned to study the changes in ultrasenic attenuation as a function of changes in grain size, size and dispersions of included matter, and other changes in metal structures that can be produced by heat treatment and working. Another phase of the attenuation studies will be an attempt to provide valid ultrasonic in- spection of welded structures. An electronic counter type of time intervalometer has been obtained which is capable of measuring the time interval between successive ultrasenic pulses to within +0.1 psec. This instrument is presently being used with an existing pulse-echo ultrasonic instrument for a very accurate meas- urement of ultrasonic velocity. |t is planned to evaluate the changes in ultrasonic velocity as a function of the parameters noted above for the ultrasonic attenuation measurements. In addition, these velocity measurements will be used in basic studies of the elastic properties of metals. Part 2 CHEMISTRY AND RADIATION DAMAGE 2.1. MATERIALS CHEMISTRY W. R. Grimes Chemistry Division PREPARATION OF CHARGE MATERIAL FOR REDUCTION TO YTTRIUM G. J. Nessle Methods have been developed for the preparation of the LiF-Mng-YF3 mixture used in the production of yttrium metal and for the delivery of the mixture into the reduction vessel. The production process consists in preparation, by a method similar to that used successfully in the past for the preparation of fluoride fuel mixtures, of a suitable mixture of LiF, MgF,, and YF, that is virtually free of oxygen, reduction of this mixture with metallic lithium to produce a low-melting yttrium-magnesium alloy, removal of the magnesium and the lithium by vacuum distillation, and consolidation of the resulting yttrium sponge to a massive ingot. The reduction, distillation, and consolidation steps are carried out by the Metallurgy Division in the manner described in Chap. 1.1 of this report. Details of the methods used in the production of the fluoride-mixture charge material by the Chemistry Division are presented below. Since conversion of Y,0, and suitable compounds of magnesium to fluorides can be accomplished on a large scale in available equipment, a pilot plant capable of supplying a quantity of fluoride mixture sufficient for the production of 200 Ib of yttrium metal per month has been designed and is being constructed. Conversion of Y203 to YF3 J. Truitt Complete conversion of an oxide to a fluoride by treatment of the solid with anhydrous HF is usually difficult because the product fluoride tends to coat the oxide and greatly diminish the rate of conversion. Since, however, HF treatment of the LiF-MgF,-YF, liquid is the final oxygen-removal stage in the yttrium production process, complete conversion of Y 0, to YF, in the first step is not essential. There- fore conditions for the treatment of Y, O, by HF have been examined in small-scale equipment similar to the 250-lb-scale equipment which was vapor used successfully for conversion of ZrCl, to ZrF . The treatment temperatures used have varied from 1100 to 1400°F ond HF flow rates have varied between 84 and 132 liters/hr for a 3-lb charge of Y203. The processing time was usually 12 hr, but one test was terminated after 6 hr. Evaluation of the product by petrographic examination, x-ray diffraction, chemical analysis for fluorine and yttrium, and by measurement of the oxygen evolved upon treatment with BrF indicated that the products of all runs were satisfactory. The conversion, as calculated from the fluorine-to-yttrium ratio, was above 97% in all cases and was, in general, above 99%. The oxygen content was 1.5% in the worst case and was usually below 0.5%. Some further testing will be required to establish optimum processing times and HF flow rates, but the results show that a temperature of 1100°F and the lowest flow rate of HF used thus far are adequate. The scale-up of the process to that to be used in the very similar 250-lb-capacity equipment should present no great difficulty. Preparation of MgF, J. Truitt Inquiries are still being made, but thus far no MgF, which passed all specifications has yet been offered by commercial sources at a reasonable price. Accordingly, conversion of MgO and Mg(OH), to MgF, by hydrofluorination of the solid materials has been investigated. The effect of temperature and HF flow rate have been examined over the same range as in the Y,0; hydrofluorination described above. The hydroxide was found to be more readily converted than the oxide and to yield an appreciably purer product. The hydroxide has the disadvantage, however, that its bulk density is lower than that of the oxide and therefore less material can be produced per run in the equipment and higher losses of powder to the gas exit train occur. Modifications have been made in the gas exit line to alleviate the latter difficulty. 53 ANP PROJECT PROGRESS REPORT A temperature of 1100°F was found to be adequate for conversion of either magnesium compound. Petrographic and x-ray examination re- vealed traces of MgO in three of four preparations made from the oxide but showed no such material in preparations made from Mg(OH),. The con- version, based on the fluorine-to-magnesium ratio obtained from chemical analyses, was above 95% in every case and was, in general, better than 98%. |t appears that sufficiently good conversion to provide feed for the liquid-phase purification step can be assured and that scale-up to 250-1b equipment will not prove to be difficult, Preparation of the LiF-MgF ,-YF, Charge Mixture J. Truitt In the present process for producing the yttrium- magnesium alloy used in the preparation of yttrium metal, a fluoride mixture containing 25 wt % LiF, 21.1 wt % MgF,, and 53.9 wt % YF, is charged to the reduction vessel. |t is essential that the fluoride mixture be very pure, especially with respect to oxide concentration, since the oxygen content of the metal depends directly on the oxide concentration of the fluoride mixture. In order for the metal to have the low oxygen content specified, the fluoride mixture must not contain more than 500 ppm oxygen. In the preparation of the charge mixture, the individual fluoride salts LiF, MgF,, and YF, are blended in the proper proportions in a 5-lb- capacity nickel reaction vessel. The vessel is purged with helium to remove entrapped air and moisture and then with anhydrous HF at room temperature. The mixture is then heated to 1500°F wunder an atmosphere of HF. At this point, the melt (mp, 1210°F) is subjected to successive treatments of hydrogen and anhydrous HF to remove the last traces of oxides and reducible impurities. At the completion of the treatment the batch is tronsferred through a sintered nickel filter and sampler into a nickel can, - which is so designed that the fluoride mixture can be transferred directly into the reduction vessel when needed. The melt is allowed to cool and the receiver can is stored on a helium header system under a positive pressure of helium. J. E. Eorgan receiver A sample of each preparation is submitted for chemical and petrographic analysis in order to 54 determine whether the purity and the composition meet the specifications. At the present time, no satisfactory method is available for determining the oxygen. content of the purified fluoride melt, but the analytical method used for determining oxygen in the YF, is being modified and adapted for use on these melts (see Chap. 2.2, this report). |f precise analyses indicate the need, steps can be taken to change the processing procedure to lower the oxygen content of the melts. Thus far, ten batches of the fluoride mixture have been made with YF, and MgF, processed from the oxides. Petrographic and x-ray onalyses have not detected oxides in these batches. The System LiF-YF, R. E. Thoma Magnesium fluoride is present in the fluoride mixture charged to the reduction equipment only because the magnesium released upon reduction drastically lowers the freezing point of metallic yttrium and permits separation of molten metal from slag at a readily attainable temperature (that is, below 1000°C). If the reduction could be accomplished at a temperature above 1600°C, the melting point of yttrium metal, the magnesium would be unnecessary, and subsequent vacuum distillation and consolidation steps could be avoided. In this instance, the binary system LiF-YF; could be used as the charge material if satisfactory compositions could be melted at temperatures sufficiently low to permit purification by the standard method (described above) in equipment of copper or nickel. A preliminary investigation of the binary system LiF-YF, has indicated that liquidus temperatures in the composition region 15 to 50 mole % YF, are much lower than those reported in the Russian literature.! Results of recent thermal-gradient quenching experiments are consistent with thermal- analysis data, to within 5°C, and show the presence of a single eutectic, containing 23 mole % YF, which melts at 699 +4°C. A previous visual observation in this laboratory had indicated a solidus temperature of 690°C in this composition region.? A single LiF-YF3 compound has been observed. Optically, it is uniaxial positive with ]E. P. Dergunov, Doklady Akad. Nauk S.5.5.R. 60, 1185 (1948). 2R. 1. Sheil, ANP Quar. Prog. Rep. Sept. 30, 1957, ORNL-2387, p 114. an ordinary index of refraction of 1.454 and an extraordinary index of refraction of 1.472. Results of quenching experiments show that this compound contains at least 40 mole % YF_. Although the data are as yet insufficient for the preparation of a phase diagram of the LiF-YF3 system, it is obvious that satisfactory compositions are available that will standard equipment. permit processing in the EXTRACTION OF LITHIUM METAL IMPURITIES WITH MOLTEN SALTS G. M. Watson J. H. Shaffer Experiments in the selective precipitation of oxides from molten fluoride mixtures have sug- gested that lithium oxide is relatively soluble in such molten salts, It should, accordingly, be possible to remove lithium oxide from molten lithium by extraction with a molten salt, and it may be possible at the same time to remove lithium nitride and lithium carbide. Since as a consequence of such a separation the lithium metal will be saturated with the extracting salt at the temperature of separation of salt and metal, it will be advantageous to use a low-melting salt mixture for the extraction in order to minimize the salt content of the lithium, Metallic lithium is such an active reductant, however, that, if contamination of the lithium by extraneous metal is to be avoided, only salts of lithium may be used. These con- siderations led to the choice of the eutectic mixture LiF-LiBr (12-88 mole %; mp, 453°C) for preliminary tests of the extraction method. Treatment of molten lithium at 600°C with the LiF-LiBr mixture should extract most of the lithium oxide from the metal, but the salt mixture will be saturated with metallic lithium. Therefore, in preliminary experiments, metallic bismuth has been used in an additional step to extract the dissolved lithium from the salt. If the lithium is completely removed, titration of an aqueous solution of the salt should provide a convenient measure of the lithium oxide removed from the molten lithium. The LiF-LiBr eutectic was prepared by adding LiF that had been purified by hydrofluorination and hydrogenation to dehydrated LiBr. The bismuth extractant was prepared by reducing commercial-grade bismuth with hydrogen at 600°C in a graphite-lined nickel reactor and transferring PERIOD ENDING MARCH 31, 1958 it through a stainless steel filter disk of 0.0008- in. porosity in a stainless steel line to a storage container; the stored bismuth was then treated with additional hydrogen at 400°C wuntil used. The lithium metal purification apparatus consists of an extraction chamber and auxiliary storage containers. The entire apparatus, including lines, is constructed of stainless steel; it is filled with a hydrogen 4tmosphere and heated to 700°C prior to use. An argon atmosphere is maintained at all times aofter the hydrogen-firing step. The purification procedure consists in trans- ferring, to the extraction chamber, a portion of LiF-LiBr salt mixture measured by differences in liquid levels in the salt storage container. The lithium to be purified is then transferred to the extraction chamber from o weighed storage container. All transfers are effected by pres- surizing with argon, and changes in transfer-line connections are made inside a small, inert- atmosphere dry box. The two-phase mixture is then agitated for 1 hr at 600°C by energizing a magnetic coil. (The optimum extraction time has not yet been determined.) Following the extraction a dip leg is lowered to the salt-metal interface and the lithium is transferred to a clean storage container. An extraction of dissolved and entrained lithium from the salt is made at 600°C by repeating the extraction procedure with liquid bismuth as the extractant. Following this, a sample of the salt, taken by means of a filter tube, is dissolved in water and its total alkalinity is determined by titration with standard acid. The alkalinity of the pure salt mixture having been previously determined, a calculation of the quantity of lithium impurities removed from the metallic lithium is then possible. One experimental determination has been completed, and the results are presented below: Weight of salt 1387 g Weight of lithium metal 675 g Salt-lithium extraction time 1 hr Salt-bismuth extraction time 34 hr Lithium impurities extracted 324 ppm {caleculated as 02) The effectiveness of these methods for purifi- cation and analysis of the purified product cannot, 55 ANP PROJECT PROGRESS REPORT of course, be evaluated on the basis of this one experiment. It is planned to test the methods by repeating the purification procedure with the same batch of lithium. If the method is effective, the amount of impurities removed as the number of extractions increases should rapidly diminish. When purification appears to have been attained, known amounts of an oxide, probably CuO, will be added to the purified lithium metal, and the process will be repeated to determine its reliability. The apparatus will then be modified to permit sampling and onalysis of the metallic lithium. PERIOD ENDING MARCH 31, 1958 2.2, ANALYTICAL CHEMISTRY J. C. White Analytical Chemistry Division DETERMINATION OF OXYGEN IN METALLIC LITHIUM A. S. Meyer, r. R. E. Feathers The precision of a method which has been proposed for the determination of lithium oxide in metallic lithium' was evaluated by analyzing samples of a dispersion of metallic lithium in paraffin. Lithium oxide was assumed to be uniformly distributed with respect to the metal in the solidified hydrocarbon, and therefore a source of standard samples of unknown, yet vniform, oxide concentration was available. The dispersions were prepared with a Waring Blendor fitted with a stainless steel mixing cup in which a helium atmosphere was maintained. In this cup, 95 g of lithium was combined with 230 g of paraffin. The lithium used had been pretreated, in an effort to reduce the concentration of dense impurities, by centrifugation of the molten metal ot a temperature slightly above its melting point. The metal was dispersed in the paraffin by heating the contents of the vessel to 230°C ond agitating it vigorously for 30 min. The dispersion was then cooled rapidly to avoid segregation of the oxide. Samples of the solidified material that contained approximately 1 g of the dispersed metal were analyzed. The analytical procedure described previously2 was followed in detail, except that the samples were wrapped with perforated silver foil to moderate the rate of dissolution of the metal. The results of the analyses are presented below: TA. S. Meyer, Jr., et al., ANP Quar. Prog. Rep. Sept. 10, 1956, ORNL-2157, p 127. 2A. S. Meyer, Jr,, and R. E. Feathers, ANP Quar. Prog. Rep. Sept. 30, 1957, ORNL-2387, p 149, 3A. S. Meyer, Jr., and G. Goldberg, ANP Quar. Prog. Rep. Sept. 30, 1957, ORNL-2387, p 150. 4, Sheft, A. F. Martin, and J. J. Katz, J. Am. Chem. Soc. 78, 1557 (1956). Lithium (g) Oxygen (ppm) 0.76 240 0.90 243 0.97 222 1.01 224 1.02 220 Average 230 Standard deviation 11 Coefficient of variation 5% Since the absolute standard deviation of the results corresponds to a titration error of only about one drop of 0.01 N standard acid, the precision is considered to be satisfactory. Further, the results indicate that the assumption that lithium oxide is uniformly distributed in the lithium-paraffin dispersion is valid. DETERMINATION OF OXYGEN IN FLUORIDE SALTS A. S. Meyer, Jr. G. Goldberg The use of the addition compound, KBrF4, which is formed by the reaction of equimolar amounts of KF and BrF,, as the active reagent for the determination of oxides in fused mixtures of fluoride salts has been described in previous reports.3 This compound, KBrF,, functions as a basic flux, and, as such, reacts readily with acidic oxides such as yttrium oxide, but it reacts slowly with basic oxides such as magnesium oxide. Sheft, Martin, and Katz? recommend the use of an acidic flux, Ber-SbFé, which is the addition compound of BrF ., and SbF ., for reactions with strongly basic oxides. Accordingly, the apparatus for the determination of oxides in fused mixtures of fluoride salts is being adapted for use with BrF2-SbF6. This flux should be ap- plicable to the determination of oxides in fused mixtures of lithium fluoride and magnesium fluo- ride. 57 A ANP PROJECT PROGRESS REPORT 2.3. RADIATION DAMAGE G. W. Keilholtz Solid State Division CREEP AND STRESS RUPTURE tested in air in order to avoid the changes of . TESTS UNDER IRRADIATION thermocouple calibration observed earlier when the ). Wilson specimens were tested in helium, and air cooling W E. Brond W. W. Davi of the specimens will be used to achieve better ise SRMRAUGS S temperature control. The earlier apparatus was N. H. Hinkle J. C, Zukas subjected to temperatures on the order of 1700°F as a result of gamma heating. Since the use of air rather than helium surrounding the apparatus would result in even more severe overheating, a “‘cold finger'’ that is cooled by circulating air is centered in each specimen (as shown in Fig. 2.3.1). MTR Experiments The tube-burst stress-rupture apparatus for tests under irradiation in the MTR was completed and final preirradiation tests are under way. This six- specimen apparatus is similar to the eight-specimen apparatus operated le-::1r|iita|',‘l except for two im- V). . Wilscr a al,, ANP Qudv. Prog. Reb. Juse 30, portant modifications. The specimens will be 1957, ORNL-2340, p 268. UNCLASSIFIED PHOTO 42859 I[||||||||||||||I]]]l||l||||l|| 3 i 2 3 INCHES T —— e Figs 2.3.1. In-Pile Stress-Rupture Specimen and Cooling Means. At the top is o complete stress-rupture specimen and capillary tube for pressurizing to the test stress. The cold finger projects from the right end. In the middle is a blind-end, cold-finger tube ready for welding into @ specimen. At the bottom is a hairpin tube for supplying cooling air to the top of the cold finger. Cooling of the specimen takes place by radiation from the inside surface of the specimen tube to the outside of the cold finger tube. At the in-pile end of the apparatus where the gamma heating is greatest, there is also some cooling by metal conduction. The partially assembled MTR appa- ratus is shown in Fig. 2.3.2. Electrical heating was used to simulate gamma heating in out-of- pile tests made to establish the design con- figuration. The MTR tests will be operated at 1500°F at stresses from 3000 to 6000 psi. The specimen tube walls are 0.03 in. thick. Out-of- pile tests of single specimens in air have given results that agree with those obtained earlier by the Metallurgy Division. Since air is to be used as the specimen environ- ment in the MTR tests, little more has been done in the study of the effects of helium and con- taminants from the experimental apparatus on the change of calibration of Chromel-Alumel and iron- constantan thermocouples.? X-ray diffraction, Curie point, and metallographic studies have been started. 2). C. Wilson et al., ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 207. PERIOD ENDING MARCH 31, 1958 LITR Experiments An examination was made of an Inconel stress- corrosion specimen with a 0,02-in.-thick wall that had operated in the LITR for 625 hr before rupture., The stress on the specimen was 2000 psi; the test temperature was 1500°F; and the specimen environment was NaF-ZrF -UF, (62.5 12,525 mole %) on the inside of the capsule and helium on the outside. The post-irradiation examination showed a maximum of 3 mils of attack at the center of the specimen where the control thermocouples were located., The attack was in the form of scattered voids at grain bound- aries. The attack on the helium side of the tube was approximately the same in terms of depth, distribution, and density of wvoids. Rupture occurred in the neighborhood of the liquid-vapor interface in the upper part of the capsule. It is possible that the temperature exceeded 1500°F in this region. In common with other in-pile stress-rupture tests, the regions of the test specimen that showed the typical severe intergranular separations of third-stage creep were localized in the neighbor- hood of the fracture. In out-of-pile tests, the UMNCL ASSIFIED PHOTO 42860 Fi Qe 21-3-2- in-pile ends of several specimens are visible. The !fi-in. tubes will supply cooling air to the cold fingers inside the specimens. Partially Assembled Stress-Rupture Apporatus Designed for Insertion in HB-3 of the MTR. The ANP PROJECT PROGRESS REPORT occurrence of the intergranular separations is more widespread throughout the specimen length, ORR Tests The apparatus-holding fixture for attaching experiment cans to the pool face of the ORR has been completed and installed, The surfaces of the reactor tank were neither plane nor well- and therefore careful shimming was to obtain the required dimensional Several sets of the experiment latch- aligned, necessary tolerances. ing mechanism and the handling tool have been designed ond built, A sample holder has also been built that will be used initially to hold thermocouples for measuring the pool face temper- ature during shakedown operations. Design work is proceeding on experiment cans and terminal facilities for experiment cables. An air-filled can has been designed and built to fit against the pool face of the reactor for use in monitoring the flux pattern in the pool facility. The can will also be used by the Operations Division to de- termine the effect of voids on reactivity. Mockups of conceptual models of in-pile creep apparatus for irradiation at the ORR have been tested in order to determine the best specimen geometry for large numbers of specimens in the presence of gamma heating that is expected to be of the order of 10 w/g. The objective is to accommodate about 18 specimens in a 15in. length of one of the lattice positions, but it may be difficult to hold the temperature down to the desired 1500°F. Instrumentation, safety circuit, and equipment layouts are being planned. RADIATION EFFECTS IN ELECTRONIC COMPONENTS J. C. Pigg C. C. Robinson 0. E. Schow The current-voltage characteristics of p-n junctions have been studied extensively. The two principal pictures currently used to explain barrier behavior are the junction theory of Shockley? and the carrier generation and recombination 1‘heory5 of Pell,* which was later extended by Sah et al, 3W. Shockley, Bell System Tech. |. 28, 435-89 (July 1949). 4E. M. Pell, J. Appl. Phy. 26, 658 (1955). 5C. T. Sah, R. N. Noyce, and W. Shockley, Proc. Inst. Radio Engrs. 45, 1228 (1957). 60 The junction theory of Shockley predicts a saturation current in the reverse direction and a forward current proportional to e7Y/%T, The carrier generation theory does not postulate saturation in the reverse direction and predicts an e?V/"kT Jependence in the forward direction. The former picture can be used to explain the current-voltage characteristics at room temperature of semiconductors having small forbidden gaps - for example, germanium —and the latter to explain these characteristics in material having large Pell has shown that even in germanium the charge-generation forbidden gaps — for example, silicon. process predominates at low temperatures., The reverse current of a junction may be ex- pressed as the sum of two terms: IR = Id + I’_g ’ where [, = the junction current, L= the generation current, The reverse junction current is given by Shockley as _ e-qV/kT) with = 1/2 1/2 IflS = Aqn (Dfl/Tfl) / where the diffusion constant, D, and lifetime, 7, are those corresponding to the location in the crystal where the carrier considered exists as a minority carrier and where V = voltage, g = electronic charge, n,p = density of electrons and holes, re- spectively, Dn,Dp = diffusion constant of electrons and holes, respectively, T Ty = lifetime of electrons and holes, re- spectively, when appearing as a mi- nority carrier, k = Boltzmann’s constant, T = absolute temperature, A = area of the barrier. The reverse current component from the charge recombination-generation mechanism as given by Pell is I =qGWA , where G is the charge regeneration rate, [ is the fraction of the space-charge region involved in charge generation, and W is the thickness is the barrier, These factors have been given in terms of the electronic properties of the material as follows: b, (1 - e—-qV/kT) G = 'rn(n + n]) + ‘rp(p + p]) ' V+e ~2c¢ g T V+VAF where n, and p, are the respective carrier densi- ties that would occur if the Fermi level were at the energy level of the recombination center, €1 is the band gap, €_is the depth of the recom- bination center from the nearest allowed band, VA is the barrier height. The barrier thickness, W (in cm), can be de- termined in terms of the barrier capacitance, C (in ppf), which for germanium is Ww=142A/C , and n,, p, by definition, are given by (¢. — € Y/kT n,=N_e T & (eg - Er)/kT . p] =Nve where N_ is the density of states in the conduction band and N, is the density of states in the valence band. A combination of these factors gives the reverse current as PERIOD ENDING MARCH 31, 1958 It can readily be seen that the I, term correspond- ing to the recombination-generation mechanism does not indicate saturation in the reverse di- rection and that this term is also more strongly dependent upon carrier lifetime than is the dif- fusion component. Ifit is assumed that the recombination-generation level is appreciably closer to one band than to the other, the expression can be simplified, If, for example, the level in question is near the conduction band, then Ty > Tb e The ex- pression then becomes ].42A2 V+ Eg —2€r I = pd T8 C V+Vag Nv (eg - er)/kT X —e (1 —e-aV/kT) | , The Irg term in I, is exponentially related to (eg_sr)/kT temperature by the factor e The I, term in I, is exponentially related to temper- -€ _/kT ature by the factor ¢ & ', It would be ex- pected, as Pell found, that the [, term predominates at high temperatures and the Irg term predominates at low temperatures in a germanium junction. Curtis et al.® found a recombination level about 0.2 ev below the conduction band, Such a level justifies the assumption made above. They found that the lifetime decreased essentially linearly with irradiation exposure, and they attributed the decrease in lifetime to the introduction of recombi- nation centers by the irradiation. which the R would be expected to increase as a result of the decreased minority carrier lifetime. Since I, = 1/71/2 and 1, 1/, L. would increase at a faster rate than I, In ad- dition, new For small-dosage irradiations in Fermi level is not changed, I recombination-generation centers 0. L. Curtis, Jr., et al., |. Appl. Phy. 28, 1161 (1957). 1.4242 gn p(V + €p = 2€r) L= C [-rn(n +n)+ 'rp(P+ Pl)] [V + VAF] (1 - e—-qV/kT) 61 ANP PROJECT PROGRESS REPORT that will add a new term lr' to I, will be intro- duced at 0.2 ev below the conduction band. If 1, is defined from the limit as V goes to zero of the slope of the reverse current-voltage characteristic, { I il li 4 (1, ) =1,+— lim — . R d q V_'Odv rg 1/2 1/2 D, D = Agp|— +Agn | — + T T p n 2 n, eg—2€r +142 —g—|—| . C s\ YAF Curves showing the effect of temperature on I, for a IN91 germanium-alloy-junction power rectifier before irradiation, after an integrated fast neutron exposure of 10'3, and after an exposure of 1074 are presented in Fig. 2.3,3. Before irradiation, UNCLASSHFIED — 4 ORNL—LR—DWG 30433 ! ~ 0,2ev | 1077 ‘ | \ A BEFORE IRRADIATION \ o INTEGRATED FAST NEUTRON FLUX OF 10" o INTEGRATED FAST NEUTRON \ FLUX OF 10" t\ 0.54 ev 10°7 \ ACTIVATION } ENERGY, O.75ev —6 I, (amp) o —8 10 340 360 380 400 420 440 460 480 103/7 (°K) Figs 2.3.3. Curves Showing Effect of Temperature on Iy for IN91 Germaonium-Alloy-Junction Power Rectifier Before and After Irradiation in ORNL Graphite Reactor. 62 the slope of the line shows an activation energy of 0.75 ev, which corresponds to € . This indicates that the Shockley diffusion theory applies to this sample. After irradiation in the animal tunnel of the ORNL Graphite Reactor to an exposure of 10'3, lo had increased by about a factor of 4. Measure- ments of barrier height showed no appreciable change, and thus no appreciable change in the Fermi level was indicated. The curve does not show a sharp break of the kind reported by Pell: in fact, the slope seems to increase at low temperatures. It is believed at present that this increase in slope is experimental error resulting from the small currents involved in measuring [ at low voltages. After an ex- posure of ]0]4, the curve assumed a slope of approximately 0.2 ev, in agreement with the ex- pectation of a predominating I _ as a result of a recombination-generation level 0.2 ev below the conduction band. The absence of saturation in the reverse current- voltage characteristics curves of germanium diodes after the units have been exposed to irradiation is thus due to a change in rectification mechanism resulting from the bombardment- produced recombination-generation centers. Such a change in mechanism will also change the forward characteristic. Hence, changes in the emitter behavior of transistors subjected to radi- ation may be expected. IRRADIATION TESTS OF MODERATOR MATERIALS FOR USE AT HIGH TEMPERATURES W. E. Browning R. P. Shields J. E. Lee, Jr. Preparation of apparatus, as previously de- scribed,” for irradiation of moderator materials at high temperatures in the ETR has been continued. The yttrium hydride capsules submitted by GE- ANPD for this experiment were examined radio- graphically and were found to have large hidden radial cracks. Additional yttrium hydride speci- mens have therefore been requested; they will be encapsulated at ORNL. The other specimens for the initial test are ready. The specimens are right cylinders, and they will be stacked in the C-33 H10 position in the w. E. Browning and R. P. Shields, ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 209. ETR in the order shown in Table 2.3.1, which is the order from the top down with the center line of the reactor between numbers 3 and 4. Table 2.3.1 also gives the specimen size, material, and gamma heating it will receive. Each capsule will contain three specimens, Table 2.3.1. Moderator Materials for Irradiation in the ETR Height of G Capsule Each 1-in.«dia Moderator amma ) Heating Number Specimen Material . (w/g) (Inv) 1 0.5 Beryllium oxide 18 2 0.75 Bery llium 23 3 0.75 Yttrium hydride 25 4 1.0 Yttrium hydride 25 5 1.0 Beryliium oxide 23 6 1.0 Yttrium hydride 18 PERIOD ENDING MARCH 31, 1958 Compatibility tests are being run to determine the effect of yttrium hydride on Inconel and on type 446 stainless steel when heated in contact, in a hydrogen atmosphere to a temperature of 900°C, The duration of the test will be several weeks. Several mockup experiments have demonstrated that the heat from the hottest specimen can be carried across a 10emil annulus filled with a helium-nitrogen mixture which will be essentially static. Also it has been demonstrated that thermo- couples can be attached to the walls of the capsule in the 10-mil annulus, Several preirradiation tests have been or are being made on all specimens. The yttrium hy- dride specimen will be studied by x-ray diffraction; it will be heat cycled; and measurements will be made of its elastic properties and dimensions. The other specimens will be studied similarly, except that they will not be examined by x-ray diffraction. The x-ray diffraction examination of yttrium hydride wili be made to determine crystal type and lattice spacing. These studies will be repeated after irradiation in order to detect changes, if any. 63 Part 3 ENGINEERING A. P. Fraas A. L. Boch Reactor Projects Division 3.1. COMPONENT DEVELOPMENT AND TESTING H. W. Savage E. Storto Reactor Projects Division LOOP FOR INVESTIGATING CORROSION IN LITHIUM-COLUMBIUM SYSTEMS D. R. Ward A forced-circulation loop has been designed, as shown in Fig. 3.1.1, for testing corrosion in lithium-columbium systems. The loop will be con- structed of columbium, and lithium will be circu- lated with the use of motor-driven electromagnets. The loop is designed to minimize the need for cladding to provide oxidation protection of welded joints. Clad joints are required only at filling- and expansion-tank connections in the coolest part of the loop. The welded columbium tubing joints in heated and cooled parts of the loop will be surrounded by a liquid metal (sodium) heat-transfer fluid or an inert gas. High-quality columbium tubing is presently becoming available only in short random lengths, and many joints may there- fore be required in the heated and cooled portions of the columbium loop. Prior to construction with columbium, the performance characteristics of the loop will be determined in an identical loop con- structed of stainless steel in which NaK will be circulated. PUMP DEVELOPMENT ¥W. F. Boudreau A. G. Grindell Irradiation Test of Bearings and Seals D. L. Gray Operation of the rotary element of an oil-lubricated centrifugal pump installed in a gamma-irradiation facility inthe MTR canal, as previously described, ! was resumed after the damaged shaft had been repaired. Examination of the shaft indicated that the lower journal bearing had been damaged and that the damage had probably occurred as a result of temporary, accidental operation of the shaft at an excessive speed. By late in the quarter, the bearing and seal parts had been subjected to a total accumulated gamma-ray dose of 1.6 x 107 rep. About 98% of the lubricant (Gulfcrest-34) is ex- ternal to the radiation field, however, and the ]D. L. Gray and W. K. Stair, ANP Quar. Prog. Rep. Sept. 30, 1957, ORNL-2387, p 32. gamma-ray dose to the bulk of the oil has been only 3 x 107 rep. Only the lubricant that remains between seal faces or has accumulated in the catch basin has received the full dose. it has been estimated that by the time the bulk of the oil has accumulated a dose of 5 x 108 rep, a 25% increase in viscosity will have occurred. The increase in the viscosity of the lubricant and the resultant bearing overheating and increased seal leakage are expected to provide information with which to evaluate the useful life of the rotary system in a radiation field. The data obtained can then be used in the design of elements and shielding with which to lengthen the useful life. The effect of irradiation on the physical properties and composition of the oil will be determined. Seal Improvements D. L. Gray Two modifications made to improve the primary face seal of a centrifugal pump for high-temperature service are being tested. @ The modifications were made to reduce the rate of accumulation of grease-like materials on the gas side of the seal where lubricant that leaks through the seal mixes with gases laden with constituents from the liquid being pumped. In one modification, the leakage of lubricant is reduced by the use of a Fulton-Sylphon seal. In the other modification, purge gas is admitted to the shaft clearance annulus between the seal and the liquid being pumped so that flow of the gas toward the liquid reduces diffusion of undesirable vapor to the seal region, The Fulton-Sylphon seal was installed ina liquid metal pump and had operated about 3700 hr by the end of the quarter at pump speeds ranging from 2400 to 3450 rpm. The rubbing speed between seal faces varied from 39.25 to 56.5 fps. Oil 3 Thus far no evidence of accumulation of deleterious deposits on the gas side of the seal has been leakage averaged less than 1 cm® per day. noted. The split-purge arrangement, described pre- viously,? had operated about 3400 hr when the test 2D. L. Gray, ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 25. 67 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 30434 MOTOR-DRIVEN \ ELECTROMAGNETS = 7~ LITHIUM EXPANSION TANK J N A — T N - HELIUM l INLET \THERMOCOUPLE LEADS i’ i‘g SODIUM EXPANSION TANK A ‘ (HELIUM ATMOSPHERE) CLAMSHELL HEATERS & A HOT-LEG LITHIUM-FILLED TUBING COOLING WATER 7 ) 1 INLET & COLD-LEG LITHIUM-FILLED TUBING — COOLING WATER OUTLET LITHIUM FILL AND x DRAIN TANK HELIUM PRESSURIZING ‘! INLET ‘ ’ SODIUM FILL AND DRAIN TANK Fi gs 3.1-]- Circulating Lithium. Diagram of Forced-Circulation Loop for Corrosion Testing of Columbium Tubing Exposed to was suspended to find the cause for increasing restriction to the flow of the gas. Deposits had accumulated in the shaft annulus and closed the gas entry ports. Since there were no deposits above the gas inlets or in any part of the seal region, it was evident that the purge gas had successfully stopped the rise of vapor from the pumped fluid. The cleanliness of the ocil-leakage catch basin and the deposits in the shaft annulus may be seen in Fig. 3.1.2, An elastomeric seal between rotating parts had hardened and it may have allowed lubricant to leak into the shaft annulus independently of the action of the pump primary face seal. Cavitation Tests P. G. Smith W. E. Thomas Experiments have shown that cavitation-induced damage of an impeller is primarily a function of the head-vs-flow characteristics of the pump. The density of the fluid being pumped affects only the pressures, and therefore pump cavitation tests with water and liquid metals will indicate the operating conditions to be expected in the pumping of lithium with similar pumps. Two tests are now under way in which pumps are being operated over a wide range of conditions in order to evaluate cavitation-induced damage as a function of time. One pump has operated under cavitation conditions for more than 4000 hr without changes in per- formance characteristics and the other has operated 500 hr. Upon completion of these tests, which are scheduled for a total of 10,000 hr, the impellers will be examined for cavitation damage. VALVES FOR USE AT HIGH TEMPERATURES . T. Dudley Satisfactory high-temperature operation of a general-purpose poppet valve with a bellows- sealed stem has been demonstrated. Valves with the plug and the seat of Stellite 6 overlaid on Inconel have been tested with NaK, and valves with Kennametal cermets® for both the plug and the seat have been tested with fused salts. Since 3The Kennometol cermet designated KM was used for the plug and the Kennametal cermet designated K-1628 was used for the seat; see ANP Quar. Prog. Rep. March 31, 1957, ORNL.-2274, p 49, for compositions of these cermets. PERIOD ENDING MARCH 31, 1958 cermets of the type tested are known to be com- patible with lithium, the valve with the cermet plug and seat would be expected to be satis- factory for use with lithium. The available valve bodies were constructed of Inconel and conse- quently could not be tested with |ithium, The valve with the Stellite 6 plug and seat was tested with NaK at 1000°F under a pressure of 50 psi. Leakage decreased from 0.75 cm3/hr to 0.2 cm3/hr after 1500 hr. The force needed to open the valve remained less than 50 Ib, and a 1300-1b force was needed to effect closure. Post- test examination showed slight creep of the seat- ing surfaces, but the closure had not been affected. The valve with the K-162B cermet seat (Rockwell A hardness, 89) and KM plug (Rockwell Ahardness, 91) operated satisfactorily when the test liquid was free of particulate matter. Particulate solids suspended in the liquid prevented full seating of the very hard, carefully lapped surfaces of the mating cermets. THERMAL STABILITY TESTS OF METAL SHELLS J. C. Amos R. L. Senn The second thin-shell test model was subjected to a creep buckling test at 1500°C in an atmosphere of helium with an external pressure of 52 psi. The shell buckled as shown in Fig. 3.1.3 in 111.15 hr. The top half of this shell where buckling occurred had previously been subjected to 339 thermal cycles in a thermal-cycling test stand. A detailed description of this thermal-cycling test was presented previously.? Two additional thin shells of the same geometry are being fabricated by welding together over-size sections and reducing the weldment to the desired final dimensions by machining. Metallurgical observation has shown that in some locations the weld metal that remains after the machining may consist of a single grain that extends across the total shell thickness. In order to investigate further the durability under thermal cycling of a vesse| fabricated as a machined weldment, one of the new shells is currently being assembled in the thermal-cycling test container. 4). C. Amos and L. H. Devlin, ANP Quar. Prog. Rep. March 31, 1957, ORNL.-2274, p 50; J. C. Amos and R. L. Semn, ANP Quar. Prog. Rep. Sept. 30, 1957, ORNL.-2387, p 45. 69 0L UNCL ASSIFIED PHOTO 30754 Figs3.1.2. Oil-Leakage Catch Basin ond Outer Wall of Shaft Annulus of Centrifugal Pump Operated with o Split=Purge Arrangement to Prevent the Rise of Vapor from the Pumped Fluid. 130d3Y S53390dd L23rodd dNV PERIOD ENDING MARCH 31, 1958 UNCL ASSIFIED PHOTO 0914 Fig. 3.1.3. Thin Shell Weldment That Buckled After 111.15 hr at 1500°F in en Atmosphere of Helium and an External Pressure of 52 psi. 71 ANP PROJECT PROGRESS REPORT 3.2, HEAT TRANSFER STUDIES H. W. Hoffman Reactor Projects Division STUDIES OF THE EFFECT OF THERMAL-STRESS CYCLING ON STRUCTURAL MATERIALS Pulse-Pump System J. J. Keyes A. |. Krakoviak Three additional tests (tests 6, 7, and 8) have been run with the high-frequency thermal-cycling pulse-pump loop! in which the fuel mixture NQF-ZrF4-UF4 (56-39-5 mole %) was the circu- lating fluid and helium was the pulsing and cover gas. The test sections were 0.485-in.-I1D, 0.415- in.-wall Inconel cylinders machined from bar stock. The operating conditions and results of tests 6 through 8 are given in Tables 3.2.1 and 3.2.2. This series of experiments has been primarily concerned with the effects of exposure time (or total number of cycles) on thermal cycling, Thus, runs 5 and 6 and runs 7 and 8 each constitute time-comparison test-pairs which were performed under conditions as nearly identical as could be maintained with long-term operation of the experi- mental equipment. In all tests the fluid flow was kept constant at 6 gpm (equivalent to 5 = 3700 Btu/hr-ft2.°F in the test section); the pulsing ]J. J. Keyes, A, |. Krakoviak, and J. E. Mott, ANP Quar. Prog. Rep. Dec. 31, 1957, ORNL-2440, p 54. frequency was 0.4 cps. The stream-temperature amplitudes were different in the two test groups. In run 6, incipient intergranular cracks were found after 22.8 hr of pulsing with a stream- temperature amplitude of +242°F. This is illus- trated in Fig. 3.2,1 which shows surface cracks to a depth of 2.5 mils. The test section of run 5, as previously reported,? had cracks as deep as 72 mils after 135 hr of pulsing. For tests 7 and 8, the stream-temperature amplitude was +139°F, with run 7 lasting 155 hr and run 8, 351 hr. The results of these tests are given by the photomicrographs of Figs. 3.2.2 and 3.2.3. The specimen used in test 7 showed variable subsurface void formation (ranging from light to heavy in concentration) to a depth of 3 mils. No cracks were found in the test section. This is to be compared with run 8 (exposed to 2% times as many thermal cycles) in which massive cracking occurred. Typical cracks which formed near the test-section entrance (region just beyond the conical inlet) are shown in Fig. 3.2.2. The cracks penetrated the 0.415-in, wall to a depth of 0.080 in. A section of l-in. sched-40 pipe ]/2 in. down- stream from the Y-section (junction of the hot and cold streams) was removed after a total 21bid., Fig. 1.7.2b, p 57. Table 3.2,1. Operating Conditions of HighFrequency Thermal-Stress 05 N - - Y Y\.\\ e = 50,000 ® % ~ = . f+2.9 2 \ x\\/}: 2200 Btu/hr - ft<-°F e 0.4 e ™~ ~— \; 0.3 ™~ ‘ i — EXPERIMENTAL RESULTS 0.2 — —] ——— ESTIMATED FROM ENTRANCE DATA FOR AIR, Npe= 50,000, INFINITE CONTRACTICN (REF 3) | l i | | I o1 q 2 3 4 FREQUENCY (cps) Fig. 3.2.7. Variation of Surface Attenuation, 7, with Frequency at x/d = 4.3, PERIOD ENDING MARCH 31, 1958 at least x/d = 1 and to study other entrance configurations of interest. THERMAL STRUCTURE OF FLUIDS FLOWING IN VOLUME-HEATED SYSTEMS Diverging Annular Channel with Screens N. D. Greene The stydy of the effect of screens in a diverging annular channel on the thermal structure of a fluid with internal heat generation has been completed. Three screens with increased porosity in the region of the outer wall were installed in the positions shown in Fig. 3.2.8, The topmost screen was of uniform porosity. A 3.8-in.-thick drilled Boltaron sheet located in the channel entrance converted the swirl entrance flow to A ORNL-LR-DWG 30439 \ /FLUID IN [ eorwaros \ "/ // I‘i“l ,J > / E /, 7/ / é 0.6 ; 77— w 7 47/ T—MEAN FLUD = / w 04 7 A - / ,, ’ ’ 0.2 L e A A -~ ”, - O ” 0 2 4 6 8 10 12 14 16 18 AXIAL DISTANCE FROM INLET (in.) Fige 3.2.12. Mean Temperature Distribution for Variable Porosity Screened Channel with One-Pump Operation. Power generation wuniform at 120 kw; Np . = 40,000+ PERIOD ENDING MARCH 31, 1958 the two-pump and one-pump conditions. This study has demonstrated that the amplitude of the surface temperature fluctuations can be markedly reduced by introducing properly designed screens into the main channel flow region. Liquid-Metal Experiment G. L. Muller® H. W. Hoffman initial studies of heat transfer in a liquid metal (mercury) with internal heat generation have been completed. @ The scope of the experiments is indicated in Table 3.2.4. To determine whether magneto-hydrodynamic effects existedin the mercury flowing through the test section due to the heavy (1000 amp) electrical current used in heating the mercury, several half-power runs were made. Table 3.2.4. Test Conditions for Volume-Heated Mercury Experiment Run Description 16 runs at flow rates ranging from Np, =6,000 to Ngp, = 150,000 2 runs at flow rates dupli- Full power Half power cating two full-power tests Low power 1 run Effect of guard heater 2 runs at different guard- on test-section tem- heater power levels perature Thermocouple calibration 4 runs at femperature levels varying fram 58°F to 222°F Electrode heat loss 2 runs at two power levels The results of these tests can be compared with the results of full-power runs at the same flow rates to establish the magnitude of this effect. Preliminary calculations have indicated that magneto-hydrodynamic corrections will be negli- gible. One low-power test was made to establish the shape of the fluid mixed-mean temperature profile along the test section. Midway in the series of full-power experiments, the steel rotor of the Moyno pump began to bind 8On assignment from Pratt & Whitney Aircraft. 79 ANP PROJECT PROGRESS REPORT within the rubber stator. The pump is a positive- displacement screw-type in which the turning rotor rubs along the stator, and the liquid being pumped provides the necessary lubrication between the rotor and stator. Because of its high surface tension relative to the usual oxide coating on steel, mercury is a poor lubricant. In order to overcome this defect, sufficient sodium metal was added to the system to produce a sodium- mercury alloy with 0.02 wt % Na. This sodium addition had a negligible effect on the thermal and electrical properties of the fluid, but it reduced the oxide layer on the steel and allowed the mercury to wet the steel. Preliminary analyses of the data indicate that the surface-to-fluid mean temperature differences 80 are greater than those predicted by the analyses of Poppendiek.” From this, it appears that the theory-vs-experiment discrepancies observed in liquid-metal wall-heat-transfer systems exist also in the volume-heated system. Since inter- facial resistances, such as gaseous films or particulate scales, do not affect the results of this study (adiabatic walls), it can be concluded that the radial conductivity (molecular and eddy) in the mercury is smaller than postulated. PO . 4. F. Poppendiek and L. D. Palmer, Forced Con- vection Heat Transfer in Pipes with Volume Heat Sources Within the Fluids, ORNL-1395 (Dec. 17, 1952). PERIOD ENDING MARCH 31, 1958 3.3. INSTRUMENTATION AND CONTROLS H. J. Metz Instrumentation and Controls Division DATA ACQUISITION SYSTEM G. H. Burger A data acquisition, or digital recording, system was purchased from the G. M. Giannini Company in order to evaluate the use of such a system in the handling or recording in digital form of such variables as temperature, flow, and pressure which are normally encountered in re- process actor instrument systems, reactor component test stands, and reactor instrument development and it is expected that, with the experience gained with the use of this equipment, a better evaluation can be made of the need for instruments and instrument systems for future reactor appli- cation. The system purchased was selected upon the basis of itsrelative simplicity and compatibility with the present use of strip-chart recorders, as well as its expected reliability (comparable with that of a Brown recorder) and its flexible design features. The system is constructed to record automatically on a typewriter and to display visually on a lamp bank the values of the millivolt inputs of ten 12-point Brown strip-chart recorders (120 points) plus a time reference. The automatic typewriter is equipped to punch tape in the Oracle code. Controls are included in the system to enable automatic testing. logging cycles tobe made at regular timed intervals in tabular form or on tape or both. The system is capable of logging 120 points in approximately 4 to 8 min, depending on the Brown recorders. Re- corders other than the Brown instruments can be used, and the recorders may be single-point, 16- point, or 24-point as well as the presently installed 12-point recorders. A self-balancing potentiometer or an equivalent device is used to convert voltage inputs to shaft position. The initial digital encoding is done by an encoding disk which is mounted on the movable shaft of the self-balancing potentiometer or other device, LIQUID-METAL-LEVEL TRANSDUCERS G. H. Burger Life tests of ORNL-designed liquid-metal-level transducers have been continued. Two of the four units being tested have accumulated 5408 hr of operation at 1200°F in NaK and the other two have accumulated 775 hr. it is planned to continue these tests indefinitely. Eight transducers were tested to investigate wetting at temperatures of 300, 400, 500, and 600°F. Seven of the transducers were regular Inconel units and one was an Inconel unit plated with iron to a thickness of approximately 0.003 to 0.005 in. The test results for all the units are now being compiled and will be included in a later report. After completion of the wetting tests, four units were installed in the test rig and the temperature was raised to 1200°F for life tests of these units. STRAIN GAGES FOR USE AT HIGH TEMPERATURES C. L. Pearce, Jr.! A search of the literature on bonded strain gages for static measurements at temperatures up to 2000°F is under way. Since the bulk of the published work on the subject is several years old and, in many cases, preliminary, some of the people who have been involved in strain gage research are being contacted to bring the survey up to date. Various groups at ORNL who have used strain gages have been interviewed; however, the use of strain gages at ORNL has been almost entirely at low temperatures. The companies that market commercial strain gages are being contacted to determine the present limits of applicability of their gages to high-temperature static strain measurements and to determine the feasibility of extending the range. Effort is being directed toward devising a rela- tively inexpensive test facility for evaluating and comparing the various types of bonded strain gages presently available. The comparisons should show wherein the high-temperature limitation occurs in each case in order to indicate the possibilities of extension of the temperature limit, A preliminary survey has indicated that the present temperature limit on reliable measurements 10n assignment from Radio Corp. of America. 81 of static strains is 900°F. Phase changes in the materials appear to be the limiting factor. TEST SET-UP FOR STUDY OF EFFECT OF RADIATION ON THERMOCOUPLES A. M. Leppert Preliminary design of an in-pile loop in which to study the effects of high-level radiation on thermocouples has been completed. In this design, sheathed thermocouples are brazed into a pot partially filled with triple-distilled sodium under high vacuum. The sodium acts as a thermal heat sink. The pot temperature is obtained by monitor- ing the vapor pressure of the sodium. A bench test is under way in order to determine the reproducibility of the vapor pressure-vs- temperature curve for sodium. Since the sodium vapor will be radioactive, it should not be brought outside the test cell to the electromanometer pressure transducer, Consequently it was decided to put an isolating diaphragm in the sodium vapor line, with a solid-filled system from the diaphragm to the electromanometer. A diaphragm and housing have been assembled and tests begun to determine effects of operating the filled part of the system and electromanometer head at varying temperatures and different fill pressures. A small pot with several sheathed thermocouples has been fabricated, and as soon as the diaphragm tests are completed the pot will " be welded to the diaphragm housing and filled. 82 Tests will be then conducted to check the re- producibility of the vapor-pressure curve. The temperature range to be investigated is 600 to 1500°F. PRESSURE TRANSMITTERS FOR USE AT HIGH TEMPERATURES C. M. Burton Two pneumatic 0- to 200-psi pressure trans- mitters with 3- to 15-psi signal outputs (Taylor Instrument Cos., Model 226R)} and Inconel bodies are now undergoing overrange pressure tests. These tests were started in an attempt to determine the aging characteristics of the instruments with respect to accuracy and to over-all safety. Cali- brations are made at intervals to detect error drift, and it is hoped that the errors will give a warning of impending failure of the primary seal. It is possible that the primary seal will deform slightly by bulging of the flat circular plates above and below the diaphragms. If so, the stresses will be somewhat relieved as a spherical shape is approached, creep at the high temperature will decrease, and the safe life of the unit will be extended, but it is possible that the accuracy of the instrument may be affected. The furnaces are shut down occasionally and micrometer measure- ments are made on the pressure transmitter bodies to detect deformation. After approximately 2500 and 1750 hr, respectively, with the two units at 1400°F and 300 psig, calibrations have not changed appreciably, ond the bodies have not deformed noticeably. Part 4 SHIELDING E. P. Blizard Applied Nuclear Physics Division . 4.1, SHIELDING THEORY F. L. Keller Applied Nuclear Physics Division A CALCULATION OF ENERGY SPECTRUM AND ANGULAR DISTRIBUTION OF GAMMA RAYS AT SURFACE OF BSF REACTOR L. A. Bowman! R. H. Lessig? W. W. Dunn? D. K. Trubey In the past it has not been possible to use directly the Monte Carlo shielding codes to check, or to predict, the experimental results of shielding experiments performed with the Bulk Shielding Facility (BSF) reactor. While theoretical calcu- lations of the energy spectrum have been per- formed3 for comparison with the experimentally measured gamma-ray spectrum,? there is only limited information on the angular distribution at the reactor surface.® The calculation reported here represents a new attempt to calculate the energy spectrum from primary gamma-ray sources and, in addition, to determine the anqular distri- bution of the gamma rays at the reactor surface. Geometry The reactor was divided into four quadrants, with each quadrant containing one quarter of the re- actor volume, as shown in Fig. 4.1,1. The thermal- UNCLASSIFIED 2—0t—059-279 // 7 7 7 / \ i’ /. y SEGOND QUADRANT NORTH FACE Fig- 4.‘-1- the Gomma~Ray Spectrum of the BSF Reactor {Loading 33). Assumed Geometry for the Calculation of neutron flux in the second quadrant was used to perform an integration of the power and the gamma- ray leakage out the center of the north face of the reactor. The integration was performed over the angles ¢ and 6 so that an angular distribution of the gamma rays could be computed. Thermal-Neutron Flux Thermal-neutron flux measurements were avail- able® for points within the volume of the second quadrant of BSR loading 33 for a reactor power of 27.36 kw. Plotting and cross-plotting these measured values produced curves which could be used to determine the flux at any point in this volume. Integration over the reactor, that is, Power = Cf qb'h(hé)dhé R eactor S L - sdpd qf’fh IQSI R? 4R sin 6 d0 d¢ , where ¢,,, = thermal-neutron flux plots), C = Ef(u)/K, Ef(u) = macroscopic uranium fission cross section = 0,0565 cm™!, K = factor to convert from fission rate to power = 3.48 x 10'3 fissions/sec-kw, yielded a power of 29,48 kw or approximately 8% more than the rated power of 27.36 kw. The (taken from the 10n assignment from Wright Air Development Center. 2On assignment from USAF. 3G. deSaussure, A Calculation of the Gamma-Ray Spectrum of the Bulk Shielding Reactor, ORNL CF.-57- 7-105 (July 31, 1957); ANP Quar. Prog. Rep. Dec. 31, 1956, ORNL.2221, p 343, esp p 347. 4e, C. Maienschein and T. A. Love, Nucleonics 12(5), 6 -8 (1954). SF. C. Maienschein, T. A. Love, and F. T. Bly, Gamma Radiation in a Divided Shield. Parts 1 and 2, ORNL-1714 (Aug. 20, 1954). 6E. B. Johnson, Power Calibration for BSR Loadine 33, ORNL CF-57-11-30 (Nov. 28, 1957). 85 ANP PROJECT PROGRESS REPORT 29,48-kw figure was used in the normalization for this calculation, since neither the integration over ¢ nor the integration over @ (power and gamma-ray leakage) accounts for the corners properly. For the calculation of the second quadrant to apply equally to the other three quad- rants, it was necessary to assume that the re- actor has flux symmetry in all quadrants. Gamma-Ray Sources The following primary gamma-ray sources were considered in this calculation: (1)} prompt fission and capture gamma rays from uranium, (2) fission- product-decay gamma rays, (3) capture and decay gamma rays from aluminum, and (4) capture gamma rays from water, These sources were grouped into six initial energy groups of 0.5 1, 2, 4, 6, and 8 Mev and were taken into account as ex- pressed in the following equation, which gives the source for a particular energy group at a point: (1) S(EG,R) = by (R) [N(pre )X () + NUfp)E. (u) + + N(ADZ (A1) + N(H)Ea(H)] ' where | S d)fh(R) = thermal-neutron flux at R, N(p+c) = number of prompt and capture gamma rays in the energy group about E0 produced in uranium per fission (see Table 4.1.1), Ef(u)zmccroscopic fission cross section of U233 (0.025 ev), N(fp) = number of fission-product gamma rays in the energy group about E_ produced per fission (see Table 4.1.1 and ref 7), N(Al) = number of aluminum capture and decay gamma rays in the energy group about E, per capture (see Table 4.1.1 and ref 7), Ea(Al)=mocroscopic absorption cross section for aluminum (0.025 ev) (see ref 8), N(H) = number of hydrogen capture gamma rays in the energy group about E, per capture (see Table 4.1.1 and ref 7), 3, (H) = macroscopic cross section of hydrogen per absorption (0.025 ev) (see ref 8). Absorption Coefficients The linear absorption coefficients, p, for the core as a function of E_ were determined by the following expression, where f; is the fractional density of the ith material: @ e (core) = ) G‘) /i _ 3) f= Ll 7H. W. Bertini et al., Basic Gamma-Ray Data for ART Heat Deposition Calculations, ORNL-2113, pp 9, 16 (duly 5, 1956). 8H. C. Claiborne and T. B. Fowler, The Calculation of Gamma Heatinec in Reactors of Rectanguloid Geometry, ORNL CF-56-7-97, p 20 (July 20, 1956). Table 4.1.,1. Equivalent Number of Gamma Rays in the Core for Each Energy Group E, Energy E N{ptc) N(fp) N(H) N(AI) Interval (Meg) per per per per (Mev) Fission* Fission* Capture** Capture** 0 - 0.75 0.5 3.31 2.972 0 0 0.75 -~ 1.5 1.0 2.55 1.864 0 0.10 1.5 - 3.0 2.0 1.687 0.891 1.115 1.53 3.0 - 5.0 4.0 0.368 0.1165 0 0.77 5.0 -7.0 6.0 0.050 0.0084 0 0.21 | 7.0 -9.0 8.0 0.0067 0.0006 0 0.35 *Taken from ref 7. **Taken from ref 8. Buildup Factors The energy buildup factors for each energy group were obtained by fitting curves between the NDA tabulated values’ for aluminum and for water, Where possible, the curve was weighted by the ratio of the electron densities of aluminum and water, Flux Computation The uncollided and collided fluxes ina particular direction at the center of the north face for each initial energy group were computed by numerically integrating the standard point kernels over the volume, V, of the differential solid angle, where S(EO,I_Q’) is the averuge source strength at [_2’ and and B(uR) is the energy buildup factor: (4a) Pyuncoltidedli®:Eq) f E,S(E,R) e "FR v 47R? dR (4b) qb’)’collide d(e'('b’EO) E, S(EO,E) e PR . =f—-— [B(uR) — 11 4R . v 4nR? The collided flux from each initial energy group was separated into effective contributions to the energy group under consideration and each lower energy group. For example, for the collided flux from the 4-Mev initial energy group: (5) ¢')’collided(E0 = 4 Mev) =q5,),(4 Mev) + - 42 Mev) + b, (1 Mev) + b, (0.5 Mov) . The contributions to each of the lower energy groups were obtained by comparing the areas in each energy group under the NDA differential energy spectrum for a point source after passage through 1 mfp (mean free path) of water for the initial energy group under consideration.? The use of the energy spectrum after passage through 1 mfp of water was assumed, since the spectrum is nearly independent of distance of penetration qH. Goldstein and ). E. Wilkins, Jr., Calculations of the Penetration of Gamma Rays, NYQ-3075 (June 30, 1954). PERIOD ENDING MARCH 31, 1958 for a number of mean free paths. The total flux in a particular direction, (;by(B,qb,E), was the sum of the uncollided flux for that energy group and the collided contributions from each higher energy group, This total flux and the uncollided flux are given in Table 4,1,2. The total and uncollided flux for @ = 0 was converted to an energy spectrum and plotted in Fig. 4.1.2 as a histogram. Also UNCLASSIFIED 2—04—058—0—204R{ PERIMENTAL 7 [CORRECTED VALUES FROM 10 NUCLEONICS 12, 5-6 {1954)] w N o 18] HIGH -ENERGY CALCULATION BY DESAUSSURE, ORNL-2221, p347 M o H PRESENT CALCULATION w HIGH-ENERGY CALCULATION BY DeSAUSSURE, ORNL CF-57-7-105 T.GAMMA RAY FLUX (phofons/cmz-sec-Mev-w-sfercdian) o 2 103 5 2 10? o 1 2 3 4 5 6 7 8 9 10 4 £, GAMMA RAY ENERGY (Mev) Fige 4412, Comparison of Calculated ond Experi- mental Spectra of Gomma Roys ot the Surface of the BSF Reactor. Polar angle 0=0 deg. 87 ANP PROJECT PROGRESS REPORT Table 4.1.2. Gamma-Ray Energy Flux as a Function of the Polar Angle, 6, the Azimuthal Angle, ¢, and the Average Energy, E Gamma-Ray Energy Flux (Mev/cm?: sec:watt:steradian) E (Mev) 0 = 0 deg 0 = 30 deg 6 = 45 deg 8 = 60 deg 6 = 90 deg Total Uncollided Total Uncollided Total Uncollided Total Uncollided Total Uncollided P = deg 8.0 2.47x 105 2.38x105 2.24x10% 2.16x105 1.92x10° 1.85x 105 1.58x10° 1.54x10° 140x10° 1.36 x 10° 6.0 3.66x10°5 3.33x105 3.230x10% 2.30x105 2.83x 105 259 x10° 2.34x10° 2.18x 105 2.08x10° 1.95x 10° 40 L5 x10% 1.39x 105 1.43x 106 1.24x10° 1.24x10° 1.09x10° 1.04x10° 9,41x10° 9.31x 105 8.48 x 10° 20 3.43x10% 270 x 10° 3.08x 10° 2,42 x 10° 272x 105 220x 108 2,37 x 10% 1.98x10% 215x10® 1.83 x 10° 1.0 2.52x 105 1.38x 106 2.26x10® 1.25x10® 2.00x10% 1,17x10° 1.75x10® 1,10x10® 1.60x10° 1.04 x 10 0.5 451 x 105 6.69x 105 3.98x 105 6.20x 105 3.46 x 105 5.94x 105 2.93x10® 575x 105 2.63x 10% 551 x 10° ¢-30 deg 8.0 247x105 2.38x 105 2.12x10° 204x105 1.92x10% 1.86x10° 1.64x105 1.5 x 105 1.34x10% 1.31x10° | 6.0 3.66x10% 7333105 3.12x 105 2.85x 105 2.84x 105 2,60 x 105 2.43x10° 2.25x10° 1,99 x 10° 1.8 x 10° | 40 1.59x10° 1.39x10°5 1.36x10% LI9x10® 1.24x10® 1,10x10® 1.08x10% 9.66x10° 8.96x10° 8.18x10° 2.0 3.43x108 270x 108 2.90x106 237x10% 2.73x10® 219x10® 2.43x10° 201x10% 2.08x10% 1.77 x 10 1.0 2.52x105 1.38x 108 2.10x10° 1.24x10® 2.01x105 1L18x10® 179x10® 1.11x10% 1.54x10® 1.02x 10° 0.5 451 x 105 6,69 x 105 3.70x 105 620 x 105 3.46 x 10° 598 x 105 3.01x 10% 578x10% 2,50x10% 545x 103 | | \ ¢ = 50 deg 8.0 2.47 x 105 2.38x 105 2.08x 105 2.01x105 1.76x105 1.71x105 1.46x10° 1.43x10° 119x10% 1,17 x 10° 6.0 3.66x10% 3.33x105 3.06x 105 2.81x105 2.60x105 241x105 217x10° 2.04x105 177x105 168 x 10° 4.0 159 x 105 1.39x 105 1.34x10® 1.18x10° 115x10% 1.04x10® 9.75x10° 8.88x 105 8.04x10° 7.45x10° 2.0 3.43x 105 270 x 105 2.94x 105 236 x 105 2.60 x 106 2,14x 10® 226 x10° 1,92x10% 191 x10® 166 x 10° 1.0 2.52x105 1.38x10% 216 x 105 1.24x10% 1.92x10® 1.17x 106 1.62x10® 1,08x10% 1.42x10% 9.83 x 10° 0.5 451 x 105 6.69x 105 3.78x10° 6.23x 105 3.27x10% 598x105 276x10% 571x105 225x10% 537 x 10° ¢ = 65 deg 8.0 2.47x 105 238x10° 205x 105 1.98x105 1.65x105 1.61x10% 136x10% 1.33x10° 112x10° 110x10° 6.0 3.66x105 333x105 3.03x10° 278x105 2.45x105 229x10° 2.02x10° 191x10° 1.66x10° 1.58x10° 40 1.59x105 1.39x10% 1.33x106 1.18x 105 1.09 x 10¢ 9.89x 105 9.16 x 10° 8,42 105 7,57 x 10° 7.07 x 10° 2.0 3.43x 105 270 x 105 2,96 x 108 2.41x 105 2.50 x 106 2.09 x 10¢ 2.15x 10® 1.85x 10%® 1.B2x 10¢ 161 x 10 1.0 2.52x10° 1.38x 105 2.15x10% 1.25x 10 1.85x10° 1,16 x10® 1,59 x 108 1.07 x 10 136 = 10® 9.64 x 10° 0.5 4.51 x 10° 6.69 x 105 3.75x 105 6.28x 105 3.12x 105 6.00x 105 2,59 x 106 571 x 105 2,09 =105 534 x 10° ¢ = 90 deg 8.0 2.47x10% 2.38x 105 2.05x105 1.98x10° 1.62x10% 1.58x10% 1,33x10° 1,30x10° 1.08x10° 107 x 10° 60 3.66x 105 3.33x105 3.03x10° 279x10% 2.39x105 2.25x10° 1.97x10° 1,87 x10° 161 x10° 1.55x10° 40 1.59x10° 1.39x108 1.33x 105 1.1Bx10C 1.07 x 104 977 x 105 8.93x10% 825x10° 7.40x10° 6.93 x 10° 20 3.43x10° 270x 105 2.96x 10 239 x 10° 3.47 x 105 2.08x 105 2,11 x10° 1.83x10® 1.79x10® 1.59x 10° 10 2.52x105 1.38x 105 2,17 x10° 1.27x 105 1.83x10% 116x 105 1.56x10% 1.06x10® 133x10% 9.60 x 10° 0.5 450 x105 6.69x 105 3.80x10° 6.35x10% 3.08x10% 6.05x10° 252x10% 573x10° 203x10® 535x 10° 88 PERIOD ENDING MARCH 31, 1958 shown are the results of calculations by Conclusions 3 . 4 deSaussure” and the experimental results, The gamma-ray energy spectrum normal to the Anaulos Distributi reactor surface, as calculated in this study, is in ngular Distribution good agreement with previous calculations by The energy flux at the surface-of the reactor, deSaussure.® The present calculation is an over- averaged over the azimuthal angle, ¢, is shown estimate, particularly in the low-energy range, in Fig. 4.1.3 as a function of the polar angle, 6. since it was assumed that the collided flux, as A typical plot of the variation of the flux with ¢ calculated by the energy buildup factors, is all when 6 = 30 deg is shown in Fig. 4.1.4. traveling in the outward direction, UNCL ASSIFIED 6 NCLA FIED "‘5’06) 2-01-059 - 28 (X1g ) 2201~ 089 - 580 £=0.5 Mev ! \\ = 4 - 5 4 T |~ £ =0.5Me ' ‘é \________“ — ° : .E £ =2 Mey : ..'5 E 1’\\4 \\ ; 2 o v \ .4 F =2 Mev _3 3 k N.E 3 M—— S . ¢ mt'n AE=1 Mey \ \\ :>: % = |y______-.- £ =1 Mev 2 2 T~ Lfi- 2 e — e} iy I ‘ < D - ) o N — 8 ——— S ! £ =4 Mev \\ Il ° @ 1 [ o ] < £ =6 Mey £ =6 Mev e : I—t— o £=BlMev *T\ $ — o £ =8 Mev 0 ‘6 20 30 40 50 60 70 80 90 0 10 20 30 40 50 60 70 80 90 8, POLAR ANGLE (deq) ¢, AZIMUTHAL ANGLE (deg) Fige 4+1:3. Total Calculated Gamma=Ray Energy Fige 4.1.4. Total Calculated Gamma-Ray Energy Flux at the Surface of the BSF Reactor as a Function Flux at the Surface of the BSF Reactor as a Function of of the Polar Angle, 6, Averaged over the Azimuthal the Azimuthal Angle, ¢, for Various Energies E When Angle, ¢, for Various Energies E. Pclar Angle 0 = 30 deg. 89 ANP PROJECT PROGRESS REPORT 4.2. LID TANK SHIELDING FACILITY W. Zobel Applied Nuclear Physics Division STUDY OF ADVANCED SHIELDING MATERIALS (GE SERIES) L. Jung The experiment designed by GE-ANP for studying the production of secondary gamma rays in con- figurations containing advanced shielding materials is being continued. The configurations used in the latest tests are listed in Table 4.2,1, and the size and composition of the various materials in- cluded in the configurations are described in Table 4.2.2. these configurations have been completed, only the While all the measurements beyond gamma-ray and fast-neutron measurements are reported here; the thermal-neutron data have not yet been evaluated. The group 2 configurations consisted of a 4-in,- thick beryllium slab, 4 in. of stainless steel as a gamma-ray shield, and a 12-in.-thick lithium hydride neutron shield with boral placed at various positions within the configurations in an attempt to reduce the number of secondary gamma rays pro- duced in the components of the shield, especially in the stainless steel. There was an unavoidable J-cm-thick oil layer adjacent to the source plate. The effect of inserting boral between the beryl- lium and the stainless steel in this group of con- figurations is shown in Fig. 4.2.1. A V-in. thick- ness of boral (configuration 2-Q) reduced the gamma-ray dose rate 100 c¢cm from the source by 34.4% of that measured in the case where no boral was used (configuration 2-A). Increasing the boral thickness to I/2 in. (configuration 2-F) reduced the dose an additional 12.2%. A further increase in the boral thickness to 1 in. (configuration 2-0) gave no additional attenuation. A comparison of measurements beyond con- figurations 2-G, 2-N, 2-0, and 2-P (Fig. 4.2.2) shows that placing all the boral immediately be- hind the beryllium was not as effective as dis- tributing it throughout the stainless steel, as long as some of the boral remained behind the beryllium. Placing 1 in. of boral immediately behind the beryllium (configuration 2-0) reduced the gamma- ray dose rate 100 cm from the source 17% below that measured for a configuration in which the boral was distributed throughout the stainless steel (configuration 2-P). However, placing '/2 in. of the boral in the stainless steel and keeping 90 another ]/2 in. behind the beryllium (configuration 2-N) reduced the dose rate at the same point by 27%. Fast-neutron dose rates beyond configurations 2-F, 2-G, 2-N, 2.0, 2-P, and 2-Q are shown in Figs. 4.2.3 and 4,2,4. No particular significance can be attached to the noticeable differences in the curves, since the reproducibility of the fast- neutron data at the LTSF is not better than 10%. Configurations 4.0 and 4-P consisted of 26 ¢m of oil adjacent to the source plate (not including the unavoidable l-cm oil gap) followed by 4 in. of stainless steel. Configuration 4-0 alsc had l/2 in. of boral immediately behind the stainless steel. As indicated by Fig. 4.2.5, there were no signifi- cant differences in the dose rates beyond these two configurations; however, a comparison of these measurements with the measurements beyond configuration 4-A, in which the 26-cm oil gap was not present, shows that the oil gap in configurations 4-0 and 4-P reduced the gamma-ray dose rate 100 cm from the source plate to 22% of that measured behind configuration 4-A. The fast-neutron dose rates (Fig. 4.2.6) showed no appreciable difference except immediately behind the stainless steel. The dose rate beyond configuration 4-P is higher by 45% than that beyond configuration 4-A at a point 44 cm from the source. At 62 cm the two curves infersect. The curves for configurations 4-A and 4-0 merge for all distances from the source >60 cm. Configurations 4-M, 4-N, 4-Q, 4-R, 4-S, and 4-T consisted of 4 in. of stainless steel and ]/4 to 1 in. of boral placed in various positions with respect Configurations 4-M and 4-N were run to determine the comparative effects of placing }/2 in. of boral in front of or behind the stainless steel. The results (Fig. 4.2.7) show that an additional reduction of 14% was obtained 100 cm from the source when the boral was placed behind the stainless steel. dose to the stainless steel. A comparison of the rates beyond configurations 4-N and 4-T shows that substituting l/4 in. of boral for ]/2 in. of boral behind the beryllium increased the gamma- ray dose rate by 13% at a distance 100 cm from the source. The measurements beyond group 4 configurations again showed that evenly distributing the boral in the stainless steel matrix, while at the same PERIOD ENDING MARCH 31, 1958 Table 4.2.1. Description of Latest Configurations Tested in GE Series? Location of Unavoidable : Configuration L . . Probe,b Oil Gap Within Description of Configuration c Number z, Configuration - (em) (cm) 2-A 4 in. of Be + 4 in. of stainless steel + 12 in. of LiH + oil 58.0 2.3 2-F 4 in. of Be + l/2 in. of boral + 4 in. of stainless steel + 59.8 2.9 12 in. of LiH + oil 2-G 4 in. of Be + ]/2 in. of boral + 4 in. of stainless steel + 61.0 2.8 ]/2 in. of boral + 12 in. of LiH + oi! 2-N 4 in., of Be + ]/2 in. of boral + 2 in, of stainless steel + ]/2 61,8 3.6 in. of boral + 2 in. of stainless steel + 12 in. of LiH + oil 20 4 in, of Be + 1 in. of boral + 4 in, of stainless steel + 12 61.8 3.7 in. of LiH + oil 2-P 4 in. of Be + 1 in. of sfalniess steel + /2 in. of boral + 2 62.0 3.9 in. of stainless steel + /2 in. of boral + 1 in. of stainless steel + 12 in. of LiH + oil 2-Q 4 in. of Be + J, in. of boral + 4 in. of stainless steel + 12 59.4 3.3 in. of LiH + oil 4-A 4 in. of stainless steel + oil 15.8 1.0 4-p4 l/2 in. of boral + 4 in. of stainless steel + oil 17.8 3.7 4-N 4 in. of stainless steel+ ), in. boral + oil 17.6 3.5 ) 4-0 26 cm of oil + 4 in. of stainless steel + ]/2 in. of boral + oil 42.6 2.0 4-P 26 cm of oil + 4 in. of stainless steel + oil 41.6 2.3 4-Q ]/2 in. of boral + 4 in, of stainless steel + ]/2 in. of boral + 19.6 3.7 oil 4-R 3 in. of stamless steel + / in. of boral + 1 in. of stainless 19.8 3.9 steel + /2 in. of boral + osl 4-5 2 in. of stamless steel + /2 in. of boral + 2 in. of stainless 19,6 3.7 steel + / in. of boral + oil 4-T 4 in, of stainless steel + ]/4 in. of boral + oil 17.6 3.7 41-A 4 in. of stainless steel (dry) + oil in an Al tank (]/s-in.-thick 15.5 1.8 (air gap} walls) 41-B 4 in. of stainless steel (dry) + ]/2 in. of boral (dry) + oil in an 16.8 1.9 (air gap) Al tank (¥ -in.-thick walls) 41-C 2 in. of stainless steel (dry)+ / in. of boral (dry)+ 2 in. of 18.0 1.7 (air gap) sfumless steel (dry) + /2 in. of boral (dry) + oil in an Al tank (/ in.~thick walls) Oil Oil only in configuration tank 3.]5d - ?Includes some configurations reported in ANP Quar, Prog. Rep. Dec. 31, 1957, ORNL-2440, p 263. Actual distance from source plate to back of solid configuration. “Includes 1-cm recession of the aluminum window in the configuration tank. dPosifion of inside surface of the aluminum window in the configuration tank. 91 ANP PROJECT PROGRESS REPORT time keeping a good thermal-neutron absorber behind the solid materials, is the most effective way to reduce the capture gamma-ray dose rate. A comparison of the dose rates beyond configuration 4-A, in which no boral was used, and 4-5, in which boral was evenly distributed in the stainless steel, shows that the boral reduced the dose rate 100 cm from the source by 51% (Fig. 4,2.8), whereas the dose rate beyond configurations 4-Q and 4-R, in which boral was used but was not placed within the stainless steel matrix, was reduced 47%. As was expected, there were no significant changes in the fast-neutron dose rates beyond the group 4 configurations (Fig. 4.2.9). Configurations 41-A, 41-B, and 41-C consisted of 4 in. of stainless steel interspersed with boral in a dry arrangement to eliminate oil gaps between the slabs. aluminum The oil was contained in a 56-in.-long tank which was placed behind the solid materials. Gamma-ray dose-rate measurements beyond these configurations (Fig. 4,2.10) show that the addition of ]/2 in. of boral behind the stainless steel reduced the dose rate approximately 46% at 100 cm from the source. A immediately further 5%reduction was realized when an additional Y in. of boral was placed within the stainless steel, but this may be within experimental error. Comparing measurements beyond configuration 41-C with those beyond a similar configuration, 4-S (in Fig. 4.2.8), shows that the dose rate 30 cm . behind configuration 41-C was approximately 61% higher than that beyond configuration 4-5. Since configuration 41-C did not have the 3.7 cm of oil distributed throughout the configuration, it is presumed that the difference in the dose rates is due to the additional inelastic scattering in con- figuration 41-C, The fast-neutron dose-rate measurements (Fig. 4.2,11) show an average reduction of 17% from configuration 41-C to 41-A caused by the addition of 1 in. of oil in place of the boral. The fast- neutron dose rate beyond configuration 4-S is approximately 35% lower than that beyond con- figuration 41-C; the comparison is based on equal oil thickness behind each configuration. This difference is due to the distributed oil in con- figuration 4-S. This series of experiments is being continued. Table 4.2.2. Physical Properties of Materials Used for Mockup Tests - of Advanced Shielding Materials (GE Series) Thickness of Material Material* per Slab Description (cm} Transformer oil Specific gravity, 0.88 at 15.6°C; nominal composition, 86.5 wt % C and 12.5 wt % H; H:C atom ratio, 1.72:1 Beryllium 10.16 48 X 49 x 4 in, metallic slabs Type 347 stainless steel 2.54 5ftx5ftx1in.slab; composition, 58.35 wt % Fe, 18.0 wt % Cr, 10.3 wt % Ni, 2.0 wt % Mn, 1.0 wt % Si, 0.08 wt % C, 0.04 wt % P, 0.03 wt % §; density, 7.8 g/cm:‘1 Lithium hydride 32,0 5x 5x 1ft slab encased in aluminum cans (k‘-in.-thick walls); composition, 95 wt % LiH and 5 wt % un- accounted for (probably oxygen and hydrogen); density, 0.75 g/cm3 Boral 1.3 48 X 48 X 0.4 in. slab with 0.05-in.-thick Al clodding; composition, 80 wt % Al, 15.7 wt % boron, 4.3 wt % carbon; density, 2.60 g/cm3 Boral 0.63 44,5 X 44,5 < 0.15 in. with 0.05-in.-thick Al cladding; composition, 80 wt % Al, 15.7 wt % boron, 4.3 wt % carbon; density, 2.60 g/Cm3 *Including cladding. PERIOD ENDING MARCH 31, 1958 QI 2—04—057— 71— 444 BORAL S.S. (in.) (in.) S 4 4 4 4 N S GAMMA -RAY TISSUE DOSE RATE (ergs/g-hr'watt) o N 5 g x 10" . T0 30 110 130 {50 170 Z4 s DISTANCE FROM SOURCE PLATE {cm) ¢ Fig. 4.2.1. Gommo-Ray Tissue Dose Rates Beyond Configurations 2-A, 2-F, 2-0, and 2-Q: Effect of Inserting Boral Behind Beryllium in Configurations Containing Stainless Steel. ANP PROJECT PROGRESS REPORT S 2—01—057—71—4145 RS 4 %X 10 N\ 4 GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr- watt) 5 Be BORAL S.S. BORAL BORAL S.S. (in.) (in,) (in.) (in.) in.) (in.) 4 4 o 4 2 Y > 4 4 4 1 Yo 10 70 a0 140 130 150 170 190 . Z,, DISTANCE FROM SOURCE PLATE (cm) Fig: 442.2. Gamma-Ray Tissue Dose Rates Beyond Configurations 2-G, 2-N, 2-0, and 2-P: Effect of Distributing Boral Throughout Configurations Containing Stainless Steel. 94 FAST-NEUTRON TISSUE DOSE RATE (ergs/g-hr+ watt) PERIOD ENDING MARCH 31, 1958 Ay s 1 2-01-057-71-416 A \é\o CONF. Be BORAL S.S. BORAL S.S. BORAL S.S. LiH NO. {in.) (in.) {in.) {in.) {in.} (in.) (in.) {in.) o 26 4 Vo 4 7 12 +0iL 2 A 2N 4 Yo 2 Yo 2 12 +O0IL | ® 20 4 1 4 12 +0IL a 2-p 4 { V2 2 Yo 1 12 +0IL 16° - 65 70 75 80 85 90 95 100 Zy» DISTANCE FROM SOURCE PLATE {cm) Fig. 4.2.3. Fast-Neutron Tissue Dose Rates Beyond Configurations 2-G, 2-N, 2.0, and 2-P, 95 ANP PROJECT PROGRESS REPORT -1 2—01—-057-71-417 10 ' FAST-NEUTRON Ti{SSUE DOSE RATE (ergs/g-hr-wott) o B CONF. Be BORAL S.S. LiH 2 NO. (in.}) (in.} (in.) {in) o 2-F 4 o 4 12 +0IL e 2-Q 4 Va 4 12+ OIL 1073 - . 65 70 75 80 85 90 95 §00 Zgs DISTANCE FROM SOURCE PLATE (cm) Fig. 4.2.4. FastNeutron Tissue Dose Rates Beyond Configurations 2-F ond 2-Q. GAMMA - RAY TISSUE DOSE RATE (ergs/g-hr-wcH) OlL (cm) S.S. (in.) 4 +0IL BORAL (in.) +0IL +0IL 26 4 2 26 4 50 70 90 ZO‘ {10 PERIOD ENDING MARCH 31, 1958 R 2-01-~057-T7i-418 130 150 DISTANCE FROM SQOURCE PLATE {(cm) Fige 4.2.5. Gamma-Ray Tissue Dose Rates Beyond Configurations 4-A, 4-0, and 4-P: Preceding Configurations Containing Stainless Steel. Effect of Oil Gap 97 ANP PROJECT PROGRESS REPORT -t - L] 2-01-057-71—-419 FAST—NEUTRON TISSUE DOSE RATE {ergs/g-hr-watt) CONF. BORAL NO. {in.) O 4-A +0IL ® 4-0 iz +0IL A 4-pP +0IL 40 50 60 70 80 90 {CO {10 2y, DISTANCE FROM SOURCE PLATE (cm) Fig. 4¢2.6. Fast-Neutron Tissue Dose Rates Beyond Configurations 4-A, 4-0, and 4-P. . 10° GAMMA -RAY TISSUE DOSE RATE (ergs/g-hr- watt) 20 40 60 BORAL 80 {in.) 1 /2 PERIOD ENDING MARCH 31, 1958 _ 2—04—057—-71-420 S.S. BORAL (in.) {in.) 4 +0IL 4 +0IL 4 Y. +oIL 4 1%, +oIL 100 120 140 160 z,,DISTANCE FROM SOURCE PLATE (cm) Fig. 4.2.7. Gamma-Ray Tissue Dose Rates Beyond Configurations 4-A, 4-M, 4-N, and 4-T: Effect of Inserting Boral in Configurations Containing Stainless Steel. 99 ANP PROJECT PROGRESS REPORT < 2—01—-057—-71— 424 GAMMA—RAY TISSUE DOSE RATE (ergs /g-* hr-watt) BORAL S.S. BCRAL S$.S. BORAL (in.) (in.) (in.) {in.) (in.) a \ Yo 4 3 s 2 ‘/2 \ 20 40 60 80 100 120 140 160 - DISTANCE FROM SOURCE PLATE (cm) Za Fig. 4.2.8. Gamma-Ray Tissue Dose Rates Beyond Configurations 4-A, 4-Q, 4-R, aond 4-S: Effect of Inserting * Boral in Configurations Containing Stainless Steel. 100 PERIOD ENDING MARCH 31, 1958 A 2-M-057-T1-422 CONF. BORAL S.S. BORAL S.S. {in)) {in.) {in.) {in.) 1/2 FAST-NEUTRON TISSUE DOSE RATE (ergs/g-hr-watt) 20 30 40 50 60 70 80 90 100 HO 120 130 140 ZO,D|STANCE FROM SOURCE PLATE SURFACE (cm) Fig. 4.2.9. Fast-Neutron Tissue Dose Rates Beyond Configurations 4-A, 4-Q, 4-R, 4-S5, and 4-T. 101 | ANP PROJECT PROGRESS REPORT e o 4 2-0t—057 —71-423 CONF S.S BORAL S.S. BORAL . NO. (in) (in.} (in.) {in.) 41-A + OIL iN Al TANK 41-8 + OIL IN Al TANK a1-C Y2 +0IL IN Al TANK NOTE: ALL SOLID COMPONENTS KEPT DRY GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr-watt) 20 40 60 80 - {00 120 140 Z5, DISTANCE FROM SOURCE PLATE {(cm) Fig. 4:2.10. Gamma-Ray Tissue Dose Rates Beyond Configurations 41-A, 41-B, and 41-C: Effect of Inserting Boral in Configurations Containing Stainless Steel. 102 PERIOD ENDING MARCH 31, 1958 | < 2-01-057-T1-424 FAST—NEUTRON TISSUE DOSE RATE (ergs /g - hr - watt) 107" 5 CONF. 5.5. BORAL BORAL NO. (in) (in.) (in.) 2 41-A 4 + OIL IN Al TANK 4-B 4 Yo + OIL IN Al TANK 10-2 M-C 2 . Yo + OILIN Al TANK NOTE: ALL SOLID COMPONENTS KEPT DRY 5 2 =3 10 . 20 30 40 50 60 70 80 90 100 Zy, DISTANCE FROM SOURCE PLATE (cm) Fig. 4:2.11. Fast-Neutron Tissue Dose Rates Beyond Configurations 41-A, 41-B, and 41-C. 103 ANP PROJECT PROGRESS REPORT 4.3. BULK SHIELDING FACILITY F. C. Maienschein Applied Nuclear Physics Division THE BSF STAINLESS STEEL-UO, REACTOR (BSR-II) E. G. Silver As previously reported,' a stainless-steel-clad UO,-fueled reactor for the BSF is being designed. This reactor, identified as the Bulk Shielding Re- actor |l (BSR-Il), is to be used alternately with the present aluminum-clad reactor (BSR-I). The utility of the BSR-II for shield test purposes derives from the nature of its gamma-ray spectrum and its geometrical configuration. A steel core construction will give rise to a gamma-ray spectrum which is harder than that from an aluminum core. Since the reactors of interest to the ANP program have, in many cases, core components of the Fe-Ni-Co system, such core materials in the test reactor will mock up the gamma-ray spectrum more successfully than an aluminum core would. The BSR-ll is designed for maximum utility as a shield-test radiation source. The core is cubical, 15 in. on each edge, and the grid plate is designed so that it does not project laterally beyond the core edge. The over-all height of the core is also minimized in order to make compact wraparound shield mockups possible. This arrange- ment will provide an intense, symmetrical, localized source of reactor radiation, Detailed descriptions of the design of the BSR-I| are included in the safeguard report for the re- actor.2 The reactor will utilize a control system of four spring-driven pairs of control plates, which can be positioned in any of the 25 fuel- element positions of the core. The acceleration of the control plates by means of springs will assure safety of the system, which, owing to its high absorption cross section, has a prompt- neutron-generation time in the core that is only about one-third as long as that of the BSR-l. Mechanical tests and analog computations carried out on the control system indicate that it is 1. s. Lewin, ANP (uar. Prog. Rep. March 31, 1957, ORNL-2274 (Part 6), p 34. E. G. Silver and 1. S. Lewin, Safeguard Report for a Stainless Steel Research Reactor for the BSF (BSR-II), ORNL-2470 {(March 12, 1958). 104 capable of safely reversing positive periods as short as 8 msec. This would be equivalent to an excess reactivity of 1,2%. However, as a safety precaution it is planned to limit the potential excess reactivity to 0.75% (prompt criticality) until a planned series of tests on the BSR-Hl and its control system at the SPERT Facility are completed. The safeguard report also contains the results of calculations carried out with respect to hazards arising outside the X-10 area from a catastrophe to the BSR-Il. It was found that the hazards were less than those which could be caused by the ORR, which adjoins the BSF, and which has been approved for operation. It is anticipated that, following approval by the Advisory Committee on Reactor Safeguards, the BSR-I| will be fabricated and placed in operation during the calendar year 1958, SPECTRA OF GAMMA RADIATION ASSOCIATED WIiTH THE FISSION OF u23s T. A, Love R. W. Peelle F. C. Maienschein W. Zobel For some time efforts have been under way at the Bulk Shielding Facility to measure the spectra of the gamma radiation which emanates from U235 (ref 3) and its products® as a result of fission induced by thermal neutrons. Information on the resulting gamma-ray distributions, given as a function of the time after fission and the photon energy, is essential for optimum design of both shielding ond heat-deposition systems for power reactors. Two experimenfcl approaches have been used thus far: in one approach, delayed- gamma-ray data were obtained by observing u23s samples that had been exposed in a reactor; in the other approach, both delayed- and prompt- gamma-ray data were obtained by observing spectra in time-correlation with pulses from a U239 _con- taining pulse ion chamber. The progress with 3F. C. Maienschein et al., Pbys. Semiann. Prog. Rep. March 20, 1955, ORNL-1879, p 51. R, w. Peelle, T. A. Love, and F. C. Maienschein, A[\2{g3 Quar. Prog. Rep. June 10, 1955, ORNL-1896, P . the experiments and the analyses for these two types of problems is discussed below. Fission-Product Gamma Rays Emitted More Than 1 sec After Fission Previous reports® have presented the results obtained from the preliminary analysis of a series of experiments in which studies were made of time and energy distributions of gamma rays given off by irradiated U?3% samples between 1 sec and 1500 sec after fission. The final analysis has been delayed by the need to develop methods® for removing from the data all nonuniqueness effects caused by the response characteristics of the multiple-crystal spectrometer used in the ex- periments. Such corrections to the previously reported values may amount to between 10 and 30%. Progress has been made, and during the coming report period it is onticipated that final analysis will be completed. The data have been partially reanalyzed to provide a better approximation to the final result. The steps which have been completed include improvements in the calculation of the number of fissions occurring in the samples used and removal from the observed pulse-height spectra of the contributions from the ‘‘tail’’ of the detector response functions for monoenergetic gamma rays. Effects from the breadth of the peaks of the response curves have not been removed but are felt to be less important. The results of this latest reanalysis are presented in Figs. 4.3.1 and 4.3.2. Included in Fig. 4.3.1 is the previously reported’ equilibrium spectrum from a rotating fuel belt, which has not been reanalyzed. Included in Fig. 4,.3.2 is a comparison with radiochemical data® of the total observed rate of photon energy release as a function of time after fission. 5W. Zobel, T. A, Love,and R. W. Peelle, ANP Quar Prog. Rep. March 10, 1956, ORNL-2061, p 250-353; W. Zobel and T. A. Love, ANP Quar. Prog. Rep. Sept. 10, 1956, ORNL-2157, p 291-96; Appl. Nuclear Phys. Ann. Rep. Sept. 10, 1956, ORNL-2081, p 95-101. Sw. Zobel et al., ANP Quar. Prog. Rep. June 30, 1957, ORNL-2340, p 294-95; Appl. Nuclear Phys. Ann. Rep. Sept. 1, 1957, ORNL-2389, p 97-98. 7R. W. Peelle, W. Zobel, and T. A. Love, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 223-26; Appl. Nuclear Phys. Ann. Rep. Sept 10, 1956, ORNL-2081, p 91-94., J. F. Perkins and R. W. King, ‘“Energy Release from Decay of the Fission Products,’”” Nuclear Sci. and Eng. (to be published); Papers Presented at the Fourth Semiannual ANP Shielding Information Meeting, Nov. 19-20, 1957, ORNL-2497, vol Il, paper II-B. ated gamma PERIOD ENDING MARCH 31, 1958 Fission-Product Gamma Rays Emitted Between 5 x 10~8 and 10~% sec After Fission During the preliminary experiments designed to aid in the determination of the energy spectrum of gamma rays emitted in coincidence with fission, counts were observed® in the region between 5 x 10~8 and 104 sec after fission. These counts are now believed to arise from delayed gamma rays. More recently, a beam of thermal neutrons from the ORNL Graphite Reactor has been utilized to obtain measurements of the intensity and energy of these gamma rays. For experimental reasons that involve random backgrounds, it was impossible to use the multiple-crystal spectrometer for these measurements. Instead, a single-crystal sodium iodide detector was vused; consequently, the results are presented in terms of pulse-height spectra rather than energy spectra. Some of the spectral data obtained at various short times oafter fission, by using standard delayed-coincidence techniques, are shown in Fig. 4.3.3. The source of the fission radiation 0.5-cm-dia by 1.0-cm-long spiral fission chamber. Time-distribution data obtained for various energy groups with the same fission source and the previously described'® time analyzer are shown in Fig. 4.3.4. Gamma rays appearing ‘‘after’” fission were compared with those ‘‘before’’ fission to avoid any effects from gamma rays coincident with fission. Absolute intensities of the gamma radiation in these two experiments are known, approximately, in terms of the intensity of thé prompt gamma radiation as observed in the same geometry and plotted in Fig. 4.3.3. (The plotted points in this case were obtained by multiplying the observed number of counts per fission per Mev by 107.) Values for the relative number of counts observed in a few pulse-height groups during the time interval after fission between 5 x 10=8 and 10~% sec are given in Table 4.3.1. Although this contribution to the fission-associ- radiation does not appear to be compared with the prompt radiation, short-delayed gamma rays are not known was a small negligible such 9F. C. Maienschein, R. W. Peelle, and T. A. Love, Appl. Nuclear Pbys. Ann, Rep. Sept. 1, 1957, ORNL- 25759, p 99~110. b ¢ IOR. W. Peelle, F. C. Maienschein, and T. A, Love, Appl. Nuclear Pbys. Ann. Rep. Sept. 1, 1957, ORNL- 2389, p 263-69. 105 ANP PROJECT PROGRESS REPORT UNCLASSIFIED UNCLASSIFIED 2-01-058-0-410 y I 3 2-01-058-0-4{! 5x10 ) 10 2 5 107" ) . : | SPECTRUM FROM CONTINUCUS BELT {USE SCALE AT RIGHT —=) 5 10—1 5 2 T o w 1072 2 - s -2 @ 5 10 + T 5 g 2 @ o » 2 T -3 ', 10 o 3 T -3 T' Q 10 L T g g 5 J— .F @ - TQOsec T2 £ = & T = - g 10 1000 sec o = E 10_4 e L Q £ 5 £ . 1550sec 5 2 2 107° 107" 5 5 2 Z e COMPTON —|— PAIR GOMPTON —|— PAIR —= 10 10-6 S 5 2 2 o’ 1077 O 1.0 2.0 3.0 4.0 50 6.0 O 1.0 2.0 3.0 4.0 5.0 6.0 ENERGY {Mev) ENERGY (Mev) Fig. 4.3.1. Gamma-Ray Energy Spectra Observed at Long Times After Fission. The errors shown are a combination of counting statistics with other uncertainties. The lines were fitted to the points by eye. All data, except for the top curve in (b), were obtained at the stated times after fission. The top curve in () is the spectrum emitted from the continuous fuel belt about 8 x 103 sec after irradiation of the belt commenced (use righthand scale). No response corrections were included in this top curve. 106 UNCLASSIFIED ORNL-LR-DWG 20558A x10 -1 T3 102 > 8 = T - c . k) ? Mev @ @ & T O5i-112Mev| ~ _, 3 §10° t 2 owr o & o c o 2 Syg -5 10°% 6 10® 1 2 5 10 20 50 100 1000 10,000 50000 TIME AFTER FISSION {sec) Fig. 4.3.2, Time Dependence of the Emission of Gomma Rays for Long Times After Fission. lower On the graph are shown the photon intensity-time distributions for six energy intervals. portion of the At the top of the graph is shown the observed integrated energy release and the compilation of J. F. Perkins and R. W. King (private communication) based on spectral studies of chemically separated fission products (use righthand scale). s pectrometer No correction for the response function was made for the photon data and only a very approximate correction was made for the energy release data. to have been observed previously. Since the energy release in a beta-ray decay would have to be larger than typical neutron binding energies if these gamma rays were to follow beta decay, the short-period radiation described above is assumed to arise from primary fission products which are formed in isomeric states. While the data given above only extend to 10~% sec, the range to 10~3 sec has been surveyed. Presently unavoidable random coincidence backgrounds were larger than the counts which might have been observed from any gamma radiation in this time region. These gamma rays follow fission by so short an interval that they may be considered *'prompt”’ PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED 7 2-0¢-058-0- 385 TIME AFTER FISSION (sec) 0425 x10 ° -sec”' Mev ™! 0.50 x 10 counts « figsion o 0.2 o4 as 08 10 12 14 16 1.8 2.0 2.2 GAMMA - RAY ENERGY (Mev) Fig. 4.3.3. Pulse-Height Spectra Observed at Short Times After Fissions The top curve represents the prompt radiations Random backgrounds have been subtracted, and lines are fitted to the points only to connect associated data. for shielding purposes. However, experiments planned to observe prompt radiation may exclude these delayed radiations. As has been reported,”’ gamma rays emitted during this period may be important in the consideration of the aftereffects of nuclear explosions.'! Prompt Gamma Rays Following a period in which extensive checks were made’ to ensure credible results, many have been obtained for of the spectrum of gamma rays measurements a new determination 's. Glasstone (ed.), The Effects of Nuclear Weapons, p 343, USAEC, Washington, D. C., 1957. 107 ANP PROJECT PROGRESS REPORT UNCLASSIFIED 2-01-058-0-386 ‘T AYERAGE ENERGY ENERGY RANGE (Mev) {Mev) 015 —0.22 0.25—0.35 0.48—092 1.48—1.42 0.19 ae Q.20 a4 0.70 a 1.20 v - ‘g8ec —1 counts - fission 0 20 40 60 80 100 (x107%) TiME {sec) Fig- 403-4' Gamma Rays Within Selected Energy Groups for Times Less than 10~ % sec After Fission. been made for random backgrounds and the prompt Time Dependence of the Emission of Corrections have gamma-ray contributions. Data have been combined for adjacent channels on the time analyzers Before combi- nation, the time-channe! width for most of the runs was (9.3 + 0.:2) x 1078 sec. given energy group correspond to independent runs. The different symbols for a The errors shown are determined from the number of recorded counts, and the lines are drawn only to connect related points. emanated in coincidence with the thermal-neutron fission of U?33, Additional experimental results will be required to allow a final evaluation of these data, but a preliminary analysis was performed in which compensation was made for the nonunique response of the spectrometer to monoenergetic gamma radiation. The results of this new experi- ment, after the approximate analysis, are shown in Fig. 4.3.5. The results are in fair agreement with those obtained in a previously reported preliminary expe.-rimenf,3 and in general agreement with those of Gamble,'? although they differ as much as a factor of 2 in some energy regions. The data shown in Fig. 4.3.5 may be integrated to yield 7.2 * 0.8 Mev/fission and 7.4 *+ 0.8 photons/fission in the energy range between 12y E. Francis and R. L. Gamble, Phys. Semiann. Prog. Rep. March 20, 1955, ORNL-1879, p 20. 108 Table 4.3.1. Fission Gamma-Ray Delay Fraction as a Function of Energy* Average Gamma-Ray Gamma-Ray Energy ND/NP** Energy Interval (Mev) {Mev) 0.]9 0!15—'0-22 0-20 i 0004 0.30 0.25-0.35 0.07 +0.03 0.70 0.48--0.92 0.038 +0.01 1.3 1.18-1.42 0.020 *0.004 Integral 0.16—1.93 0.057 +0.003 *These data and the quoted errors apply to count-rate data as obtained in pulse-height groups from a small single crystal, Interpretation of these results as true photon inténsities might lead to errors of a factor of 2 or more. **ND = total number of counts per fission observed in the specified energy intervol for delay times between 5 % 10~ 8 and 10— sec; Np = number of counts per fis- sion associated with prompt gamma rays in the same energy interval, 0.3 and 10 Mev. The error estimates in Fig. 4.3.5 and in the integral quantities are preliminary, since an accurate error analysis is impossible until the experiment has been concluded. It is expected that future experiments will increase the accuracy of the data for energies 1.5 Mev. be performed to yield data between about 80 and 800 kev. The results of a thorough analysis of the existing and anticipated data will be made available as soon as possible. above An additional experiment may TOTAL-ABSORPTION GAMMA-RAY SPECTROSCOPY G. T. Chapman T. A, Love A series of gamma-ray spectral measurements important for ANP shielding design is planned for both reactors at the BSF, the aluminum-clad BSR-l and the stainless-steel-clad BSR-Il. A new spectrometer system, called the BSR model 1V gamma-ray spectrometer,'3 which is now under 13R. W. Peelle, T. A. Love, and F. C. Maienschein, ANP Quar. Prog. Rep. March 31, 1957, ORNL-2274 (Part 6), p 27. 20 -1 - Mev —1 photons - fission PERIOD ENDING MARCH 31, 1958 UNCLASSIFIED 2-01-058-0-388 ¢ COMPTON SPECTROMETER o PAIR SPECTROMETER MEASURED ENERGY RESOLUTION (FULL PEAK WIDTH AT HALF HEIGHT} i 2 PAIR SPECTROMETER — = 2.0 3.0 GAMMA ~ Fig- 4.3.5. shown were obtained from counting statistics, and the over which the results were averaged. The resoclution horizontal barss The line is drawn only to connect the Energy Spectrum of Gamma Rays Observed Within 10~7 sec After Fission. —— COMPTON SPECTROMETER T Average photons - fission -Mev 1 from 7.7 to 10.5 Mev 4.0 RAY ENERGY (Mev) 5.0 6.0 7.0 The ordinate errors energy errors represent in each case the energy interval functions of the spectrometers used are indicated by the pointss This plot represents a preliminary analysis of the dota, and systemotic errors as large as 15% may occur in some ener regions. ’ Yy g g g construction, will be used for these measurements, and it is hoped that the detector for the system will be a total Such a crystal interpretation of complex gamma-ray spectra owing to the increased probability that a gamma ray may lose nearly all of its energy in the crystal. This, effect, would simplify the observed pulse-height distribution for any given gamma-ray absorption’’ Nal(TIl) crystal. potentially may facilitate the in energy by eliminating a major portion of those lower-energy pulses arising from scattered photons escaping from the crystal. Currently, multiple-crystal spectrom- eters'4 are used for obtaining reactor spectral data. Unfortunately, these spectrometers are various MF. C. Maienschein, Multiple-Crystal Gamma-Ray Spectrometer, ORNL-1142 (July 3, 1952). 109 ANP PROJECT PROGRESS REPORT complicated electronically and have low effi- ciencies which require time-consuming operations. The total-absorption spectrometer, with its greater sensitivity, should simplify spectral measure- ments through the possible elimination of the complicated two-speed coincidence and pulse- height analysis system required with the muitiple- crystal spectrometers, The total-absorption crystal used in the model 1V spectrometer system will probably be a 9-in.-dia right cylindrical crystal of Nal(Tl), one of which is currently being investigated ot the BSF.'3 A %-in.-dia by 2-in.-deep well was drilled into the truncated end of the crystal (Fig. 4.3.6) in order to reduce the effects on the spectral response of Compton scattering near the surface where the gamma radiation enters the crystal. The crystal has been investigated by introducing gamma rays of energies ranging from 0.279 Mev 3g. T, Chapman, ANP Quar. Prog. Rep. June 30, 1957, ORNL-2340, p 295; G. T. Chapman and T. A. Love, ANP Quar. Prog. Rep. March 31, 1957, ORNL.- 2274 {(Part 6), p 25. MOUNTING RING NEOPRENE REFLECTOR SUPPORT 3/4-in.-DIA x 2-in-DEEP WELL (ngoa) to 2.76 Mev (Na?4) into the truncated end of the crystal through long lead collimators. The response of the crystal to the two gamma rays from Na?4 before the well was drilled into the crystal is shown in Fig. 4.3.7. The pulse- height distribution was obtained with a 20-channel pulse-height analyzer. The distortion that appears on the low-energy side of the 2.76-Mev peak was noticeable when any gamma ray of energy greater than about 1.6 Mev was introduced into the crystal. It was postulated that this distortion was due to the interaction of gamma rays near the surface of the crystal and to subsequent escape of the scattered gamma rays from the small-diameter region of the cone. The response of the crystal to the same gamma- ray energies with the same conditions, but with the gamma rays collimated into the bottom of the well, is shown in Fig. 4.3.8. The shape of the distribution curve indicates an improvement over the distribution shown in Fig. 4.3.7 in that the 2.76-Mev peak is more symmetrical about the UNCLASSIFIED 2-01-058-0-203 _—FOR EQUIPMENT ATTACHMENT \\\\\\m N — ! S N i iy %/\ . l {/ ¢ Fig. 4.3.8. Sketch of 914-in.-dia Sodium lodide (Thallium-Activated) Gamma-Ray Spectrometer Crystal Showing :z-in.-diu by 2=in«~Deep Well in Truncated End. 110 UNCLASSIFIED 2-01-058-0-164 300 I F s 137 Mev 280 —— ——d o —— 260 U O I . v ] 240 : : e e 220 Ll . Lor i, C e e R NOTE: DASHED CURVE = 200 L J1! sHows PossiBLE CoMPTON | E SCATTERING CONTRIBUTION., 5 > 180 , : { L e < E 5 160 |- : e e @ < w140 bW 1 276 Mev '_ ® a { o . Q 420 | - : - = % » 3 100 [ - H-— - -4 Q & 80 ! RESOLUTION ™R T | 1% —of V| t- ] 60 AU f . ll,, A 7 ;oo » 40 —— 7 i ,,,15,,,,,, ll‘ ‘. 7 i \ 20 P k__ .‘{ o LA . o \ \ I nt' \ 0 b Vereee 0 10 20 30 40 50 60 70 80 90 RELATIVE PULSE HEIGHT Fig. 4.3.7. Response of the 9/-|n.-d|o Crystal to 2.76- and 1.37-Mev Na24 Gamma Ruys Collimated into the Truncoted End of the Crystal. The distortion on the low-energy side of the 2.76-Mev peak was postulated to be the result of multiple interactions in the small-diameter region of the cone. peak channel. A nearly constant resolution of about 15% was observed for all energies during these tests. This effect has not been explained as yet. In an attempt to improve the percentage resolu- tion, seven 3-in.-dia photomultiplier tubes were mounted on the crystal. A voltage divider was constructed to allow an accurate adjustment of the individual photomultiplier tube gains and data were obtained with the 256-channel analyzer. The response to the Na?4 gamma rays collimated into the well, as before, is shown in Fig. 4.3.9. The shape of the curve indicates an improvement PERIOD ENDING MARCH 31, 1958 UNCL ASSIFIED ORNL—LR—DWG 23045 ] i 70 ; ‘ 6o L 1 J‘4.37 Mev ‘ 50 ’/ ] 40 1 S 30 —- ‘L, - g — oo 10 s * =g J a A g COUNTING RATE (ARBITRARY UNITS) ! L oo T ke e 0 20 40 60 PULSE HEIGHT Fig. 4.3.8. Response of the 9/-tn.-d|u Sodium lodide Crystal to Na?4 Gamma Rays Callimated into the Bottom of the Well and Observed with Three 5-in.-dia Photomultiplier Tubes. The ratio of the peak area to the tatal area is 0.73 for the 1.37-Mev gamma ray and 0.67 for the 2.76-Mev gamma ray. in that the distribution is not so flat near the peak channels. It is believed that the addition of the well improved the response. (The new 256-channel pulse-height analyzer has only recently been put into operation, It is necessary, because of the variable dead time of the analyzer, to install an automatic dead-time clock to facilitate the accurate subtraction of the back- ground from the data. This equipment is being installed in the analyzer.) A comparison of the photofractions, that is, the ratio of the area under the full-energy peak to the area under the total distribution curve, ob- tained with the well in the crystal with those calculated by Berger and Doggett'® is presented in Fig. 4.3.10. The calculations were made by using a Monte Carlo technique for a 5-in.-dia by 9-in.-long cylindrical crystal. The measured photofractions agree to within about 3.5% with the calculated values. Since in both cases the gamma rays were in the energy region below ]6M J. Berger and J. Doggett, |. Research Natl. Bur. Standards 56, 355 (1956). 11 ANP PROJECT PROGRESS REPORT 3 Mev, and since the radial loss is low for crystals as large as 5 in. in diameter, it is believed that the comparison is probably realistic, regardless of the difference in crystal geometry. A com- parison with other existing spectrometers is pre- sented in Table 4.3.2. This table shows the UNCLASSIFIED 2-01-058-0-387 4000 1.37 Mev 3600 3200 3 2800 4 2400 P 4 | 2.76 Mev 2000 - o COUNTING RATE (arbitrary units) 1600 ——- X 1200 ® . | \ 4 800 - % @ '0'\'\'1& . 400 ag? k‘ofi‘l’:‘. i k 0 %, 0 20 40 60 80 100 120 140 RELATIVE PULSE HEIGHT Fig. 4.3.9. Response of the 9/ in.~-dia Sodium lodide to Na24 Gamma Ruys Collimated Bottom of the Well and Observed with Seven 3«in.-dia Photomultiplier Tubes. Crystal into the Mo measurement of the photo- fraction was attempted since the equipment did not allow an accurate background subtraction. measured photofractions for the total-absorption spectrometer and those measured with multicrystal spectrometers. '4 The values for the multicrystal spectrometers were obtained from a rough fit to data taken by Peelle, Love, and Zobel.17 The total-absorption spectrometer, with the present crystal, does not compete favorably with the other spectrometers in terms of resolution, The efficiency of the total-absorption spectrometer is, however, two to three orders of magnitude greater than the efficiencies of the multiple- crystal spectrometers, 7W. Zobel, private communication. UNCLASSIFIED ORNL-LR-DWG 23046 0.16 CZ) /D = 3 045 - o ) wl : | 044 0.9 \ T EASURED WITH 9 ¥g-in.~dio CRYSTAL 0.8 |- ! ~ do 'Lu — w 7 e ' T CALCULATED FOR S5-in,—dig x 9—in. 0.6 CRYSTAL, NBS 56,335 (1956) — 0.5 ‘ O 0.4 0.8 1.2 1.6 2.0 2.9 2.8 £, (Mev) Fig. 4.3.10. Ratio of Peak Area to Total Areq, p(E )s for the 9/-|n.-d|c| Sodium lodide Crystal as a Funchon of the Energy of Gamma Rays Collimated into the Well. Table 4.3.2, Photofractions Obtained with Several Spectrometers for Gamma Rays of Various Energies Photofractions Gamma-Ray Energy Total-Absorption Pair Compton (Mev) Spectrometer Spectrometer Spectrometer 0.662 0.87 0.81 0.903 0.74 0.76 1.37 0.73 0.99 0.67 1.85 0.69 0.94 0.57 2.76 0.67 0.84 0.38 112 A STUDY OF RADIATION HEATING IN SHIELDS K. M. Henry experimental work with high-efficiency has been devoted almost entirely to Past shields measurements of the radiation attenuation of the shields, with little investigation of the con- comitant heating produced in the shields by the radiation. Since it may be necessary to provide cooling channels in aircraft shields, a study of the heat production in a large slab mockup of a typical shield was undertaken at the BSF. This work is now in progress, with the objectives of the experiment being (1) to obtain experimental information that can be used to verify the results of the best available methods for calculating the energy deposition in a shield, (2) to obtain data that can be used directly in current shielding problems, and (3) to obtain flux and dose measure- ments within a shield so that correlations can be made between these measurements and heat deposition. The shield mockup currently being investigated 2-01-058-0-391 OIL TANK; 3/g-in.-THICK ALUMINUM WALLS SLABS 4 ft SQUARE WATER - s % 5 ft SQUARE 3‘/2in. I g =|a 2@ =T I &5 3 1k 3% - INSTRUMENT K{ 67 HOLES 1, 2, & 3 IS BT Tl waTer Ffi§-—HEATING SAMPLES BULK L SHIELDING REACTOR c £ - (LOADING NO. 33) {| [T [ <] & - sin.—a] b= GULF 10-C INSULATING OIL GAMMA SHIELDING MATERIALS % in. OF IRON . ) —_— A — 3in. OF LEAD 2 in. OF MALLORY 1000 Plon View of BSF Shield Heating Figo 4.3.]]. Experiment with Heating Samples in the Fore Position. PERIOD ENDING MARCH 31, 1958 consists of a 4-in.-thick slab of beryllium, fol- lowed by a gamma-ray shield and 16 in. of lithium hydride. The lithium hydride is divided into two sections of 4 and 12 in., respectively, as shown in Fig. 4.3.11. varies in composition, consists of 3 in. of lead, 3 in. of iron, or 2 in. of Mallory 1000. The entire shield is submerged in an oil-filled aluminum tank positioned against the Bulk Shielding Re- actor, with the horizontal axes of the shield and reactor coincident. Measurements of the heating within the shield caused by radiation from the reactor are obtained by attaching thermocouples on or near thin disk- type samples located within the shield. The heat loss, or cooling of the shield, is also observed after the reactor is removed from the vicinity of the shield. The heating-measurement samples, which are The gamma-ray section, which each 1.25 in. in diameter, are encased in con- tainers made of the same material (see Fig. 4.3.12). The samples themselves are surrounded e i UNCLASSIFIED 2—01—-058—0- 392 0.037-in.-0D CERAMIC SUPPORT TUBES, 0.009-in. THICK WALLS B 1.25-in, A DIAMETER EXTERNAL VIEW SECTION A-A FROM EDGE - 2in. i - CERAMIC FEED-THROUGH INSULATOR 1M FOR THERMOCOQUPLE LEADS {(SEALED IN WITH "DEVCON B") i ® LOCATION OF THERMOCOUPLES - SAMPLES MATERIAL THICKNESS r (@ IRON 0.2 in. 1 e LEAD 0.4 in. D & £l £l MALLORY 1000 0.1 in. / / ~NF 712 BERYLLIUM 0.4 in. O O % vrw,,L TWO HALVES OF CONTAINER CEMENTED TOGETHER WITH "DEVCON B PLASTIC STEEL" SECTION B-B Details of Metallic Heating Somples, Locations for BSF Figo 4-3.]2. Containers, and Thermocouple Heating Experiment. 113 ANP PROJECT PROGRESS REPORT by layers of air except at points where support tubes are attached. (Because of fabrication difficulties, it is necessary to encase the lithium hydride sample in a cadium-covered aluminum case.) Each sample case contains four thermo- couples, one attached to each face of the sample and one attached to each inside face of the sample case. The cases are plugged into recessions in the faces of the shield sections along the shield axis. In a specific shield assembly all the samples are either on the sides of the shield sections facing the reactor (fore position) or on the sides away from the reactor (aft position). The diameters of the shield cases used in the aft position are larger than the diameters of those used in the fore position so that radiation streaming through crevices straight through the shield can be avoided. At the interfaces of the shield sections, vertical wells extend downward through the shield to the samples. Flux and dose AR 6 2-01-058-0-393 7= TIME, p=POWER (ARBITRARY) DETECTOR: U235 FiSSION CHAMBER HOLE { {Be—Fe INTERFACE) HOLE 2 (Fe—LiH INTERFACE) THERMAL—NEUTRON FLUX (neutrons-cm =2+ 1. p71) HOLE 3 {LiH—-LIH INTERFACE) o 4 8 12 16 20 24 DISTANCE ABOVE SHIELD AXIS (in) Figl 4-3-]3- Interfaces os a Function of Distance Above the Shield Axis (BSF Shield Heating Experiment). Thermol-Nevtron Fluxes at the Shield 114 rate measurements are made by instruments positioned within these wells (U?3% fission detector, Hurst dosimeter,and ionization chambers). The wells also accommodate the thermocouple leads. All planned measurements with configurations in which lead and iron were used as the gamma-ray shield have been completed. In addition, measure- ments with configurations using Mallory 1000 have been completed for the samples in the fore position. The work has been delayed temporarily in order to permit the Mallory 1000 activity to decay sufficiently so that the shield can be approached. The thermal HOLE 3 (LiH—LiH INTERFACE) g ._ o z 5 l._ w o 2 0 4 8 12 16 20 24 DISTANCE ABOVE SHIELD AXIS (in} Fig. 4.3.14. Fost-Neutron Dose Rates ot the Shield Interfaces as a Function of Distance Above the Shield Axis (BSF Shield Heating Experiment). SGRapHITE * / -1 GAMMA-RAY DOSE RATE (ergs - Fig- 4-3-]5. PERIOD ENDING MARCH 31, 1958 A 2-01-058-0-395 f=7TIME; p=POWER (ARBITRARY} DETECTOR: IONIZATION CHAMBER HOLE { (Be—Fe INTERFACE) HOLE 2 (Fe—LiH INTERFACE) HOLE 3 (LiH-LiH INTERFACE ) 4 8 12 16 DISTANCE ABOVE SHIELD AXIS (in) 20 24 Gamma-Ray Dose Rates at the Shield Interfaces as o Function of Distance Above the Shield Axis (BSF Shield Heating Experiment). 2-01-058-0-396 * | | l-—— REACTOR AT CONSTANT POWER ——-——J r/REACTOR SCRAMMED {HEAT SOURCE REMOVED) 30 || | 32 | f 30 { A A ,... o f |\\y—-8e SAMPLE TEMPERATURE (TWO RUNS) o & Be CASE TEMPERATURE (TWC RUNS) 20 || 7 % : R | g | / * | )/U | 26 ! , P | | .7 | 24 lfi { 22 g : L | 0 1:00 2:00 300 4:00 5:00 6:00 7:00 TIME (hr} Figu 4-30]6. Shield Heating Experiment). Temperature of the Beryllium Sample at the Be«Tank Wall Interface as a Function of Time (BSF 115 ANP PROJECT PROGRESS REPORT L 2-01-058-0-397 36 REACTOR AT CONSTANT POWER — REACTOR SCRAMMED (HEAT SOURCE REMOVED) 34 ' ’ A ‘ | 3} | \\/‘/—Fe SAMPLE TElMPERATURE (TWOiRUNS) %0 ] Ty 5 [ ‘ 3 Fe CASE TEMPERATURE (TWO RUNS) w 5 | \ E 28 o < [ 3 % s | , P /4 [ | \ | 26 | 7, | | \q\c\ / ! ER\D:: "1 | T 22 ; —~ [ 20 — " ‘ 0 e ® 00 2:00 3:00 4:00 5:00 6:00 7:00 TIME {hr) Figs 4.3.17. Temperature of the Iron Sample at the Be-Fe Interface as a Function of Time (BSF Shield Heating Experiment). 2-01 -058—!—398 30 | | | REACTOR SCRAMMED 28 (HEAT SOURCE REMOVED) |——wm REACTOR AT CONSTANT POWER *—_—l ‘ 0 l /LiH SPECIMEN TEMPERATURE (TWO RUNS) L 26 b ; = D P : | @x Ll e Z 24 | o I Cd —COVERED Al CASE TEMPERATURE I l { TWO RUNS) i ol | 0 1:00 2:00 3:00 4.00 5:00 6:00 7:00 TIME (hr) Figs 4.3.18. Temperature of the Lithium Hydride Sample at the Fe-LiH Interface as a Function of Time (BSF Shield Heating Experiment). 116 PERIOD ENDING MARCH 31, 1958 4.4. TOWER SHIELDING FACILITY L C. E. Clifford Applied Nuclear Physics Division AIRCRAFT SHIELD TEST REACTOR EXPERIMENT AT THE TOWER SHIELDING FACILITY Y. R. Cain J. L. Hull B. J. Workman' F. J. Muckenthaler In cooperation with Convair, Fort Worth, an experiment is now being conducted at the Tower Shielding Facility in which neutron and gamma- ray dose rates and thermal-neutron fluxes are being obtained under conditions similar to those encountered with the Nuclear Test Airplane (NTA) program but in the absence of the airplane structure. The NTA program at Convair included shielding measurements taken inside and outside the NTA, ]Convuir, Ft. Worth. both while in flight and while on the ground. It also included measurements in the absence of the airplane with the Aircraft Shield Test Re- actor (ASTR) on a cradle 12.5 ft above a concrete pad. All these measurements will be compared with measurements taken in air at the TSF to determine magnitudes of ground and aircraft- structure effects. |t will also be possible to obtain a more accurate source term for the ASTR in this experiment, since ground effects can be greatly reduced. The ASTR consists of an array of MTR-type fuel elements mounted horizontally between two grid plates. Demineralized water is used as moderator, reflector, and coolant. The fuel element arrange- ment can be seen in Fig. 4.4.1, which also shows UNCLASSIFIED ORNL -LR-DWG 28942 . GAMMA MONITOR . AFT GRID . PRESSURE VESSEL LEAD CTASE . CONTROL ROD . MAIN SHIELD TANKS U W ) - FUEL ELEMENTS FORWARD GRID LEAD RINGS AND DISKS FORWARD SHIELD TANKS ION CHAMBER Som~e Fig. 4.4.1. Cutaway View of the Aircraft Shield Test Reactor (ASTR). 117 ANP PROJECT PROGRESS REPORT the location of the various reactor shield tanks, which can be filled or drained remotely, and Fig. 4.4.2 shows the locations of the ASTR, the NTA crew compartment (also shown in Fig. 4.4.3), the cylindrical crew-shield mockup (also shown in Fig. 4.4,4), and various detectors in the positions used in the NTA, The TSF experiments are being conducted in three phases: (1) data are being obtained for comparison with data taken at Fort Worth with the reactor center line 12.5 ft above ground level; (2) the spectra of neutrons from the ASTR are being determined, and (3) measurements are being made for comparison with NTA detector station measurements. The configuration used in Phase 1 is shown in Figs. 4.4.5 and 4.4.6. The reactor support, which is attached to the hoist cables, allows the re- actor to rotate 180 deg. The support truss for the detector and the crew compartment or crew- shield mockup is suspended from two hoist cables at one end and is attached to the reactor support at the opposite end. In Phase 1, groups of one dosimeter, one anthracene scintil- lation counter, one bare BF3 counter and one cadmium-covered BF3 counter are suspended from the truss on the reactor center line at dis- tances of 33, 59, and 100 ft from the reactor center. The 100-ft detector station is shown in Fig. 4.4.7. Measurements are made with the various detectors as a function of reactor angular orientation, with the center line of the system 12,5 ft above the concrete apron, as a function of altitude, and as a function of reactor angular orientation at full altitude. fast-neutron A neutron collimator assembly will be suspended from the truss in Phase 2. Neutron emulsions 118 ORNL-LR-DWG 28943 PLAN VIEW ® GAMMA AND NEUTRON DETECTORS, DIRECT BEAM SHIELDS 4 GAMMA AND NEUTRON DETECTORS, NO DIRECT BEAM SHIELDS CYLINGRICAL CREW-SHIELD MOCKUP SHADOW SHIELD ASTR — e ——a— INSTRUMENT CAPSULE CREW COMPARTMENT SIDE ELEVATION NOTE: DATA TAKEN AT SiIX POSITIONS INSIDE THE CREW COMPARTMENT Fige 4.4s2. Detector, Crew Compartment, and Crew- Shield Mockup Positions in the NTA. will be placed within the collimator to determine the fast-neutron spectra emitted from the ASTR with the side tanks emptied. In Phase 3, the NTA crew compartment and the cylindrical crew-shield mockup will be suspended from the truss at positions duplicating those found in the NTA. Measurements with all types of de- tectors will be made at all the positions used in the NTA, as shown in Fig. 4.4.2, except those in the tail section of the airplane. PERIOD ENDING MARCH 31, 1958 UNCL ASSIFIED PHOTO 43314 Figs 444.3. The NTA Crew Compartment. 119 ANP PROJECT PROGRESS REPORT UNCL ASSIFIED PHOTO 43315 " 120 4.4. The Cylindrical Crew-Shield Mockup Used in the NTA, 4. Figl UNCL ASSIFIED PHOTO 43318 BEER L CEr E bl ol PERIOD ENDING MARCH 31, 1958 B o S N e U,.MUJUNM..@ |HU_HL,,,,,.HH < [IsoEl gt -1I...|\_.1|.|1\ — I Ml e R e o .rrr...rrlllu Il-.r. e o n...llnq._.l e e i g = e PS .r||...|...|1l||r_.1ln.l|.b.l., .Irl.]...ll.p.... .Ii-.. P?..I.l_pr .n..:. {thlr..L— - WY N Y w v T @ £ - o < = - a ) = o o w 2 v o = vl < o = = W i - 5 . ANP PROJECT PROGRESS REPORT UNCLASSIFIED PHOTO 43146 T TS T\ e:.-_ ha ..-gr I TIR—NNTR 7 4“ _fi'?i'lu s .I'-___.* Y ’lh ”.“nli H; g”“.';' *- w A TN m\m (7 N o """""M ..;‘::l “_l‘ e — ——— e 122 Figs 4¢4.6. Closeup of the ASTR ond Two Groups of Detectors Suspended from the Truss. PERIOD ENDING MARCH 31, 1958 N\ uncLassiFIED) M= PHOTO 43317 | Fig. 4.4.7. Detector Station 100 ft from the ASTR. 123 OVOONOOGE WN — & e & & & e & & e 4w NN NN MNMNAN — — ed od od o md =d =md NN B W = O VO ON B WA — . e . . . . . « e . . . . . . 27. 28. 29, 30. 31. 32. 33. 34. 35. 36. 37. 38. 39. 40. 41, 42, 43. 44-45, 70&—&—:‘1106‘)5>>Im§—lnm>smx’s&—:br‘or';—o'nig—n;on>-n:orno£nm-nos— W, Allen . 5. Billington . F. Blankenship . P. Blizard . J. Borkowski . F. Boudreau . E. Boyd . J. Breeding . B. Briggs . R, Bruce . D. Callihan . E. Center (K-25) . A, Charpie . E. Clifford H. Coobs . B. Cottrell . L. Culler . R. Cuneo H. DeVan . M. Doney . A, Douglas . B. Emlet (K-25) . P. Fraas H. Frye . T. Furgerson « J. Gray . L. Greenstreet . R. Grimes . G. Grindell « Guth . S. Harrill . Hikido . R. Hill . E. Hoffman . W. Hoffman . Hollaender . 9. Householder . H. Jordan . W. Keilholtz . P. Keim . L. Keller . 1. Kelley J. Keyes A, Lane . S. Livingston f‘.:rl ) - INTERNAL DISTRIBUTION 47. 48. 49, 50. S1. 32, 53. . 35, 56. 57. 58. 59. 60. 61. 62. 63. 64. 65. 66. 67. 68. 69. 70. 71. 72. 73. 74, 75. 76. 77. 78. 79. 80. 81. 82. 83. 84. 85. 86. 87-89. 90-96. 97. 98-100. ORNL-2517 C-85 — Reactors-Aircraft Nuclear Propulsion Systems . Lyon . MacPherson . Maienschein . Manly . Mann . Meghreblian . Miller . Morgan . Murphy « Murray (Y-12) . Nelson . Nessle . Overholser atriarca Penny . Perry Pigg . Reyling . Savage . Savolainen . Schultheiss Scott . Shipley imon . Skinner . Snell orto . Susano . Swartout . Taylor O-sr N U@Iumorgiigpg;v Erm-ppITC . Trauger . Trubey . Watson . Weinberg White . Wigner (consultant) . Williams . Wilson . Winters W Zobel ORNL - Y-12 Technical Library, Document Reference Section Laboratory Records Department Laboratory Records, ORNL R.C. Central Research Library DEEmME-PEOEmME-Omp 125 EXTERNAL DISTRIBUTION 101-103. Nyr Force Ballistic Missile Divigin s 104-105. ANPR, Boeing, Seattle 106. AFRR, Boeing, Wichita 107. AFPN, Curtiss-Wright, Clifton * 108. AFPR\Douglas, Long Beach § 109-111. AFPR, Dguglas, Santa Monicg 112. AFPR, Lo\kheed, Burbank 113-114. AFPR, Lockgeed, Marietta 115. AFPR, North\merican, Ca 116. AFPR, North Agerican, Dfiiiney 117. AFPR, North Angrican, LIl Angeles 118. Air Force Special{Yeapofif Center 119. Air Research and Dy veidinent Command (RDTAPS) 120. Air Research and Déke lment Command (RDZPSP) 121. Air Technical Intellidyfille Center 122. Air University Library 38 123-125. ANP Project Office, § 126. Argonne National LabilRory 127. Armed Forces Speciaiille®ons Project, Sandia 128. Armed Forces Specialeafens Project, Washington 129. Army Ballistic MissijilAgeRcy 130. Assistant Secretary diDeferge, R&D (WSEG) 131. Assistant Secretary he AlQForce, R&D 132-137. Atomic Energy Comniil#sion, Wishington 138. Bettis Plant (WAPD 139. Bureau of Aeronauti ha Park vair, Fort Worth 126 140. Bureav of Aeronautfill General Rggpresentative 141. BAR, Aerojet-Genejlilk, Azusa 142. BAR, Convair, Sanliego 143. BAR, Goodyear Aiglaft, Akron 144, BAR, Grumman Aigllaft, Bethpage 145. BAR, Martin, Baltfllore 146. Bureau of Ships 147. Bureau of Yards afill Docks 148-149. Chicago Operationfi@Office 150, Chicago Patent Grijlp 151. Curtiss-Wright Corfliration 152-155. General Electric (Jnpany (ANPD) 156. Hartford Area Offig 157. Jet Propulsion Lallratory 158. Knolls Atomic Poyilir Laboratory 159. Lockland Area Off e 160. Los Alamos Scienffic Laboratory 161. Marquardt Aircraftffll_ompany 162. National Advisoryl_ommittee for Aeronautics tCleveland , 163. National Advisory ommittee for Aeronautics, ashington 164. Naval Air Develop ent Center 165. Naval Air Material Center . 166. Naval Air Turbine Test Station 167. 168. 169. 170. 171. 172-175. 176. 177. 178. 179. 180. 181. 182. 183. 184-185. 186-203. 204-218. 219. fon, Livermore bn Medicine rnia Radiation Laboratory, Livermore ent Center Service Extension, Oak Ridge and Development, hission, Oak Ridge Operations 127 128 ORNL-528 ORNL-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 ORNL-1375 ORNL-1439 ORNL-1515 ORNL-1556 ORNL-1609 ORNL-1649 ORNL-1692 ORNL-1729 ORNL-1771 ORNL-1816 ORNL-1864 ORNL-1896 ORNL-1947 ORNL-2012 ORNL-2061 ORNL-2106 ORNL-2157 ORNL-2221 ORNL-2274 ORNL-2340 ORNL-2387 ORNL-2440 . Reports previously issued in this series are as follows: Period Ending November 30, 1949 Period Ending February 28, 1950 Period Ending May 31, 1950 Period Ending August 31, 1950 Period Ending December 10, 1950 Period Ending March 10, 1951 Period Ending June 10, 1951 Period Ending September 10, 1951 Period Ending December 10, 1951 Period Ending March 10, 1952 Period Ending June 10, 1952 Period Ending September 10, 1952 Period Ending December 10, 1952 Period Ending March 10, 1953 Period Ending June 10, 1953 Period Ending September 10, 1953 Period Ending December 10, 1953 Period Ending March 10, 1954 Period Ending June 10, 1954 Period Ending September 10, 1954 Period Ending December 10, 1954 Period Ending March 10, 1955 Period Ending June 10, 1955 Period Ending September 10, 1955 Period Ending December 10, 1955 Period Ending March 10, 1956 Period Ending June 10, 1956 Period Ending September 10, 1956 Period Ending December 31, 1956 Period Ending March 31, 1957 Period Ending June 30, 1957 Period Ending September 30, 1957 Period Ending December 31, 1957