ORNL-2095 . C-85 :‘Reactors—.Aircmft Nucliear Propulsion Systems This decument consists of 198 pages. Cop;&g?#‘of 300 copies. Series A, AIRCRAFT REACTOR ENGINEERING DIVISION DESIGN REPORT ON THE AIRCRAFT REACTOR TEST A. P. Fraas and A. W, Savolainen May 1956 DATE ISSUED OAK RIDGE NATIONAL LABORATORY Operated by UNION CARBIDE NUCLEAR COMP ARY A Division of Union Corbide and Carben Corporation Post Office Box X Oak Ridge, Tennessee RESTRICTED DATA This decument contains Ré'st'f‘i'iyz;‘éefi“‘-putq as defined in the Atomic Energy Act of 1954, [ts tronsniittal or the disclosure of its contents s . . - ; ol s in any manner to an unauthgrized person is prohibited. » ’ RO . . W@ NN W . A, Abattieilo . G. Affel C. Amos . B. Bachulis . J. Barton . E. Bealle . Bender . W, Bertini . S. Billington . F. Blankenship . P. Blizard . L. Bech . 4. Borkowski . F. Boudrecu . E. Boyd . A. Bredig . 4. Breeding . B. Briggs . E. Brewning . R. Bruce . B. Callihan . W. Cardwell . 5. Carlsmith B C S M H D F E A C W G M E R W F A D R L. P. Carpenter C. R R C W J. 4. W D G S. R F C R S J L D E W L E. Center (K-25) . A, Charpie . B. Clarke . E. Clifford . G, Cobb A. Conlin H, Coobs . B. Cottretl . D. Cowen . AL Cristy J. Cromer . S, Crouse . L. Culler . W. Cunningham . Curry . M. DeCamp . H. DeVan . M. Doney . A. Douglas . R, Dytke . K. Eister . B. Emiet {K-25) C-85 — Reactors—Aircroft Nuclear Propulsion Sysiems INTERNAL DISTRIBUTION ORNL--2095 47. 48, 49. 50. 51. 52. 53. 54. 35. 56. 57. S8. 59. 60. 61. 62. 63. 64, é5. 56. 67. 68. 69. 70. 71 72. 73. 74. 75. 76. 77. 78. 79. 80. 81. 82. 83. 84. 835. 86. 87. 88. 89. 90. 21 92, W. K. Ergen D. E. Ferguson . Foster . Fraas Frye . Furgerson . Gray Gray . Greenstreet . Grimes . Grindell . Heestand . Helton . E. Hoffman . W. Hoffman . Hollaender ., Holland . Householder Howe . Hudson Jordan . Keilheltz . Keim . Keller . Kelley . Kertesz FOAOr-—r—4 I ~“rmru=EITrdew . Lees . Lindauer . Livingston . Lyon . Manly . Mann Mann . McDonaid . McPherson . McQuilkin . Meghreblian . Milford . Miller . Moore . Morgan CHPAANAEMEMNAAAMCEIMENINOOEASP CPIMIDPEIEDOES P> & OMe-T3 # = I'=s 0w OH e T PO - X Ll < ) i ] W m.__E - T 4 24 w ; W I i T > i o =2 “FILLER PLATES 5, e Y S S e e R3L . ONTROL. C L N ~ " Lo ISLAND R SNy [T ~. TANK - Sl oY _ SRz Fal ST Z o 3 . S BE LA Z : @ o o ¥ == >l > prd = & Ll - \ a ul = = = ' TR, e il = s SeSEETE ww 1 [ P fod o 1 o e ul . el w 35 v Ul TR 1 [ o Z i \ C4 ; a ot o ) i L o = o O ] = w e o2t . a4 - Q ow o 2oL 2 ) o : . T o ! =2 o ] a2 o7 i T T uw = &y u I_ = T c & Ul &= e ¢ = - rlJ_ & W - L. & iy . L h i e < et = e ical Section Through Reactor Assembly. Vert Fig. 1. 10 TABLE 1. ART DESIGN DATA Power Design heat output (kw) Core heat flux Core power density (max/avq) Power density, maximum (kw per liter of core) Specific power (kw per kg of fissicnable material in core) Power generated in reflector (kw) Power generated in island (kw) Power generated in pressure shell (kw) Power generated in lead layer (kw) Power generated in water layer (kw) Materials Fuel Reactor structure Moderator Reflector Shield Primory coclant Reflector coolant © Secondary eoclant Fuel System Properties Uranium enrichment (% UZBS) Critical mass {kg of U23%) Total uranium inventory (kg of U235) Consumption at maximum power (g/day) Design lifetime (hr) Design time at maximum power (hr) - Burnup in 500 hr at maximum power (%) Fuel volume in core (Ffa) Total fuel volume (Fts) Meutron Flux Density in Core 2-=:lsen:"'v') 10% ev <2 77 7 UEIE CONMOCENTRATION IMOLE %) o o SEERET ORNL-LR-DWG 1177 REFLECTOR THICKNESS =30 CM REFLECTOR THICKNESS =40 CM Fig. 5. Effect of Reactor Dimensions on Concentration of check consistency with theory and the funda- menfal constants, The first assembly was a basic reflector-moderated reactor with two regions — fuel and reflector. The fuel region contained uranium and a fluorocarbon plastic, Teflon, to simulate the fluoride fuels, and the reflector region contained beryllium. The fuel region was constructed in the shape of a rhombocuboctahedron (essentially, a cube with the edges and corners cut away to give U235 in Fuel. cctagenal cross sections) to approximate a sphere within the limitations imposed by the shape of the available beryllium. The system was made critical with 24.35 [b of U235, The assembly was then modified to include three regions with a beryllium island separated from the reflector by the fuel. The first of the three.region assemblies had no Inconel core shells, the second included Inconel core shells without end duels, and 19 EXTRAPOLATED REFLECTOR THICKNESS = 30 CM FOTAL YEID fWESTMENT (LB} EXTERNAL FUEL VOLUME = 4 FT° EEERET ORNL~LR-DWG 1176 EXTERNAL FUEL VOLUME = 8 FT3 Fig. 6. Effect of Reactor Dimensions and External Fuel Volume on Total U235 fpvestment. the final one mocked up the reactor, including the end ducts. The results of these experiments are presented in Table 4 and in reports on the indi- vidual experiments (see "‘Bibliography’’). A tow-nuclear-power, high-tempercture critical ex- petiment was also performed.? The reactor section 4A. D. Calithan et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 58. 20 of the assembly closely resembled the current de- sign of the ART in that it included the annular fuel region separated from the beryllium island and reflector by Efg-in.ufhick inconel core shells of the proper shape. Photographs of the assembly are shown in Figs. 9, 10, and 11. The mockup differed from the ART principally in that the fuel was not circulated and there was no sedium in the reflector- moderator regions. The system was filled initially, SHECRET ! HNL-t R DWG EXTRAPGLATFD REFLECTCR CHNL DWG 1196 THICKNESS — 20 ¢m POWESRR DENSITY RATIO y / T B 7 0 o §;§ — T K 0 D g p o N e s . . - -~ g o o ’74‘;&,/7 B e o5 - e e >y iy - < 23 e, e @ e s TS < g | Fig. 7. Effect of Reactor Dimensions on Outside Peak-to-Average Power-Density Ratic in Core of Reflector-Moderated Reactor. SRoRET AR ] R~ WG 47 EXTRAPOLATED REFLEGTOR ORNL-[.R=-DWG 't THICKNESS = 30 cm [eed <2 CER CENT THERMAL FISSIONS <& - "~ N 3 - ) 2y s . iz, - s o . fs;":;f;p & P 5> e L Fig. 8. Effect of Reacter Dimensions on Per- centage of Fissions Caused by Thermal Neutrons. for cleaning ond testing, with a 50-50 mole % mixture of molten NaF and ZrF,. Increments of molten Na,UF, (with the uranium enriched to 93% U233) were then added to the NaF-ZrF , mixture in the sump tank. After each addition of Na UF ,, the mixfure was pressurized into the core and then draired. This procedure was continued until the critical concentration was attained, The control safety rod was located within a 1.50-in.-ID Inconel thimble along the vertical axis Fig. 9. Particlly Assembled lsiand Showing the Lower Half of the inconel Core Shell and the Upper Half of the Beryllium Reflector for the High- Temperature Critical Assembly. of the beryllium island. The rod was a cylirdrical annulus of a neutron-absorber compact with @ den- sity of 6.5 g/cm3, The principal coastituents of the compact were Sm,0, (63.8 wi %) and (5d, 0, (26.3 wt %); the outside diameter of the absorber section was 1.28 in., and the gnnulus was % in. wide. The system operated isothermally at 1200°F, normally, but with the electrical heaters available the temperature could be raised to 1350°F. The critical fuel concentration was found to be 6.30 wi % {2.87 mole %) uranium, and the excess reactivity was about 0.13% Ak/k. The over-ali temperature coefficient of reactivity between 1150 and 1350°F was shown to be negative and tc have a value of 2 x 105 (AL/k)/°F. An increase in the uranium concentration of the fuel from 6.30 to 6.88 wt % resulted in an increase in reactivity of 1.3% Ak/k. The control rod had a value of 1.7% Ak/k when inserted tc o point 4 in. above the 2} TABLE 4. COMPOSITIONS AND DIMENSIONS OF REFLECTOR-MODERATED-REACTOR CRITICAL ASSEMBLIES Three-Region Assembly with Core Shells of Three-Region Assembly with }’S-in.-Thick k|6,'°ifi"’7"'|’ii‘:k }fls'i""ThECk }é‘i"'"ThiCk inconel Core Shells end End Ducts Aluminum laconel Inconel Assembly number CA-200 CA-20b CA-2c CA-211 CA-21-2 CA-22 CA-23 Beryllium island Volume, 2 0.37 0.37 0.37 1.27 1.27 1.27 1.75 Avercge radius, in. Spherizal section 5.18 5.18 5.18 5.18 5.18 5.18 7.19 End ducts 3.86 3.86 3.86 3.86 Mass, kg 19.4 19.4 19.4 67.0 67.0 67.9 92.2 Fuel region (excluding shells and interface plates) Volume, £ 1.78 178 1.72 2.06 2.06 247 1.56 liters 50.4 50,4 48.8 58.3 58.3 76.0 44,3 Average radius, in. Sphericol section Inside £.24 5.24 5.31 5.31 5.31 5.31 7.32 Outside 2.51 9.51 2.44 9.44 9.44 9.44 9.44 End ducis inside 3.99 3.99 3.99 3.99 Outside 5.28 5.28 679 5.28 Distance between fuel sheets, in. 0.639 0.284 0,142 0.142 0.173 0.142 a.142 Mess of components Teflon, kg 99.38 99.27 94,37 108.88 108.18 126,77 81.54 Urarium loading, kg 5.00 11.74 22.07 26.02 21.57 30.45 19.97 U235 oading, kg £.66 16.94 20.56 24,24 20.67 28.35 18.62 Uranium densify,b g/cn13 0.446 0.370 0.435 0.451 235 densi?y,b g/cm3 0.092 0.217 0.421 0.416 0.345 0.405 0.420 Urenium coating material, kg 0.05 g.11 0.20 0,25 0.2} 0.30 0.1% Scotch tape, kg 0.11 0.11 0.1 0.15 0.15 0.15 Q.15 Core shells ond interfece piates Mass ot components, kg Aluminum 5.85 1.10 .10 1.10 L1 1.10 1.10 inconel g 13.68 27.73 53.02 53.92 54.62 58.26 Refiector Valume, f13 22,22 22,22 22,22 20.88 20.88 20.45 20.88 Minimum thickness, in. 11.5 115 11.5 1.5 115 115 1.5 Mess of components, kg Beryliium 1155.0 1155.0 1155.0 1094.1 1094. i W077.1 1094.1 Aluminum 29.2 29,2 29.2 29.2 29.2 29.2 29.2 Excess resctivity os loaded, % 0.9 0.3 0.4 ~3 0.14 ~3 9.19 Experimental criticat mass, kg 4.35 10.8 19.8 19tz 19.9 24172 18.4 of U235 . “Only one end duct wes enlarged. bMass per unit velume of fuel region. “Mass required for a critical system with the poison reds removed, 22 Fig. 10. Outer Core Shell and Peartially Assembled Beryllium Reflector Critical Assembly. UNCL ASSIFIED § PHOTO 24437 | of the High-Temperature 23 Fig. 11. Completed High-Tempercture Critical Fuel Reactor. 24 UNCLASSIFIED PHOTQ 2440 Assembly of the Reflector-Moderated Circuloting- mid-plane, An analysis® of the experimental re- sults and the differences between the critical as- sembly ond the ART indicates that the critical concentration of the ART will be between 4.6 and 5.4 mole % uranium, theat is, well within the {imits of sclubility in the fuel mixture to be used in the ART. REFLECTOR-MODERATOR The design of the reflector-moderator region pre- sented several problems. Heat will be generated in the reflector by the absorption of gamma rays coming from the fuel and heat exchanger regions and by the siowing down of fast fission neutrons, Gamma rays will also result from parasitic capture of neutrons in the structural material and the cool- ant. One particularly strong source of hard gamma rays will be the Inconel shell that separates the fuel ennulus from the outer reflector. These gamma rays will be absorbed over an appreciable volume because the photon energy will be high and the attenuation rather small, A lesser amount of heating will also result from the generation of gamma rays by neutron capture in the beryllium. The heat generated by radiation in the various regions of the reactor and the heat transfer from the fuel system to the reflector and island cooling circuits are given in Table 5, SA. M. Perry, ANP Quar. Prog. Rep. Sept. 10, 1953, CRNL-1947, p 33. The cooling system designed for removing the heat from the beryllium in both the island and the reflector and from the Inconel shells is illustrated in Fig. 1. There are 120 cooling passages in the island and 288 in the refiecter; these passages are 0.187 in. in diameter. The dimensions of the cocling system and the flow characteristics of the coclant, sodium, are given in Tables 1-3. Sedium was chosen as the coclant because of its excellent heat transfer properties and reasonably low neutron- capture cross section. Experimental evidence has established the fea- sibility of operation of a sodium-beryllium-Inconel system if the temperoture of the system is main- tained below 1250°F, Cycling tests have also shown that the thermal stresses that will be set up in the beryllium should not give serious trouble. FUEL SYSTEMS Core Hydrodynamics The hydrodynamic characteristics of the reactor core are intimately related to those of the pumps, because the pumps must be placed in the lowest temperature portion of the circuit, that is, just ahead of the core inlet. This is necessary partly because of the fairly high stresses in the impellers and partly because of the shaft seal problem, which is discussed in the following section on ‘‘Pumps TABLE 5. HEAT TO BE REMOVED BY REFLECTOR AND ISLAND COOLING CIRCUITS Heat to Reflector Cooling Circuit Heat to Isiand Cooling Circuit {(Mw) (Mw) Radiation heating Beryllium 1.52 0.82 Reflector B4C tile 0.48 Pressure-shell B4C tile 0.01 Control rod 0.18 Filler plotes, south head C.03 Pressure shell 0.12 Reflector outer inconel shell 0.15 North-head liner 0.03 Transfer heating Through island core shelt 0.87 Through reflector core shell 1.73 From fuel-to-NaK heat exchanger 0.16 0.10 Total 4.04 2.16 25 and Expansion Tank." All the initial fayouts em- pleyed axial-flow pumps coaxial with the istand, but because the structural problems associated with the long impeller overhang proved to be diffi- cult, the hydrodynamically less desirable arrange- ment employing centrifugal pumps was chosen as the more practicable solution, Further it seemed fikely that two pumps either in series or parallel would be required if boiling of the fuel as a conse- quence of afterheat was to be avoided in the event that one pump failed. Preliminary analyses of various centrifugal pump—~ core inlet configurations indicated that the high- velocity streams frem the pump impelier would be likely to change completely the flow pattern ob- tained within the core. High-velocity streams are disinclined to diffuse once they have become separated from the wails of the pump volute, and their momentum con carry them all the woy from the impeiler through the plenum chamber, through the vanes at the core inlet, and even through the core itself to such an extent that flow separation and, sometimes, flow reversal in the core would tend to occur., The most promising configuration gppears to be that in which the pump volutes discharge tangentially into a cylindrical extension of the core inlet, as in Fig, 2. The high swirl velocity induced in this region gives a system relatively insensitive to the stoppage of one pump. From the shielding stendpoint an ideal circu- lating-fuel reactor would have very tiny inlet and outlet ducts to minimize both neutron leckage through these ducts and fissioning in regiens close to the outer surface of the reflector. While the relations between end-duct size and shield weight are very complex, there is o strong incentive to minimize the end-duct size. This, coupled with hydredynamic considerations associated with the pump design, particuiarly the velute-discharge areaq, and reactor physics studies, led to the choice of the core inlet proportions and hence the basic core fayout of Fig. 12, The core of the reactor consists of a divergent- convergent annular passage that is symmetrical about the equatorial plane; that is, the converging and diverging sections have the same shape. The area perpendicular to the flow path ot the equator is approximately four times the area dt the inlet or discharge ends. The equivalent cone angle of the divergent and convergent sections is aporoximately 28 deg (included angle), 26 In the design of the core it has been considered necessary, in order to avoid local regions of ex- cessive temperature, to effect a uniform circum- ferential distribution of the flow around the infet annulus and to assure that the fuel will traverse the divergent section of the core without incurring stagnation or reversal of the fluid boundary layers, Furthermore, it is desired that the regquired flow conditions obtain with only one pump in operation so that the reactor can be run at some appreciakle fraction of rated power with one of the two fuel oumps inoperable, It is well known that the flow of fluids in diver- gent channels is subject tc the growth and the eventual separation and reversal of boundary layers. The orobability of the occurrence of these phe- nomenc increases as the degree of divergence, as represented by the inlet-to-cutlet area ratio, and the rate of divergence, as represented by the equi- valent cone angle, are increased. Flow in a diver- gent channel is also adversely affected by uneven distribution either circumferentially or radially ot the inlet. Common industrial practice calls for divergence rates to be of the order of 7 to 10 deg included angle when preceded by several diameters straight run of pipe. Where care has been taken to achieve a symmetrical velocity profile with thin boundary layers at the inlet, nonreversed flow has been obtained in divergent channels having a 20-deg equivalent cone angle with an area ratio of 2:1. Both the degree and the rate of divergence of the ART fuel annulus are greater than those oreviously encountered, even under special test conditions. In addition, the necessity for com- pactness of the reactor rules out the possibility of achieving even distribution and thin boundary fayers at the infet by the most common methed, namely, @ convergence of the channel immediately ahead of the inlet preceded by several diameters straight run of pipe. ' Fortunately, it is not necessarily essenticl that the axial velocity distribution ccross the fuel annulus be uniform; the essential requirement is freedomn from hot spots, paorticularly at the walls. The wall hot spot, or boundary layer heating effect, is very much o function of the intensity of eddy diffusivity, or turbulence, in the core. Thus, from the standpoint of boundary layer heating, a non- uniform velocity distribution might actually be more desirable if the turbulence level were high than would a more uniform velocity distribution if the turbulence level were fow. DT ORNL-L.R-DWG 16043 - ¥oymi | O | Lo < —_—— K2 W0 : ~ il o S o e M/l_ Lo — 3 0 e ““““““““““““ ey~ 3400 40 dOL DISTANCE FROM EQUATOR {in.) Fig. 12. Core Layout. 27 The practice of interposing screens across a flow channel to achieve symmetry of the velocity profile is widely followed, It is also well known that screens ocross the discharge of a divergent chan- nel tend to stabilize flow and to delay stognation or flow reversals. In the ART fuel annulus it was highly desirable to avoid the use of screens for either purpose, since they would require increased pumping power to overcome the added resistance to flow and since they would, if used in the fissiening zone, increase the critical fuel concentration be- cause of their poisoning effect. The core flow problem was therefore twofcld. Good flow distribution was to be obtained at the inlet, in the limited space availchle, with either one pump or both pumps coperating; and flow through the core was to be without stagnation or other effects that would give local hot spots in the fiuid. Also, these conditions were to be achieved with o minimum of pressure loss and, preferably, without insertion of Inconel bedies into the fissioning zone of the core. The desired flow distribution at the inlet was obtained with a swirl-type heoder, which provides space at the inlet for circulation of an excess of fuel; that is, the volume of fuel which circulates around the inlet is greater than that which enters the inlet, and thus even circumferential distribution of the flow results. The approximate volume of the core was dictated by fuel concentration considerations, and core annulus radii were set at the irlet and ot the equator, A simple two-dimensicnal flow analysis indicated that pressure gradients resulting from passage curvature would be small relative to those resulting from divergence, An arbitrary cosine curve connecting the mean radii at the inlet and equator was therefore used as the passcge mean line, The shape of the annulus was then deter- mined by superimposing on the mean line a sched- ule of cross-sectional areas such that the resulting static pressure gradient at any peint weuld be a function of the locel dynamic pressure, Yon Doenhoff and Tetervin® found the function H, in the expression dH ‘6 dg\ /2 ' 6_;__ _ ea(H—b)l: ~ (= L@’)(fi) —o(H — ar)J . X \g dx/\T, / 6A. E. von Doenhoff and N. Tetervin, Determination of General Relations fjor the Bebhavior of Turbulent Boundary Layers, NACA<772 {1943}. 28 to be a criterion for boundary layer separation, where a, b, ¢, and d are constants, If is the shope parameter, g is the local velocity pressure outside the boundary layer, x is the distence along the axis of the channel, 0 is the boundary layer momentum thickness, Ty is the wall friction. tf H, 0, and the term 2¢/7, are assumed to be con- stant with x, the shape parometer con be written where A and B are constants, P is static pressure, and (1/4} dP/dx can be used as a criterion of separation. Obviously (1/q) dP/dx is minimized if it is constant throughout the divergence. However, if this were the case, the passage weuld still be diverging at the equator, and an undesirable dis- continuity would result in the shape of the core shells af this point., The divergence is therefore modified so that it is higher at the inlet than would be the case for P dpP - = a consfant, g dx and if is zero af the equator, The core shape thus cobtained was tested for flow profiles, with water as the working fluid and with a simulated swirl-type header. The results showed that the tangential velocities were high throughout the core but that a small upward velocity component prevailed clong the inner wall of the diverging section. At all other points the axial velocity was downward. The reason for this flow pattern is obvious. Rotation of the fluid about the axis of the cere creates a radial pressure gradient. Decay of the rotational velocity component as @ result of friction as the fluid moves axially creates an axial pressure gradient that is positive along the inner wall and negative along the outer. These gradients are algebraically additive to the positive gradient resulting from the divergence of the passage. The net effect favors flow reversal on the inner wall. It wos determined by test that if a solid-body- rotation pattern (retational velocity compenent di- rectly sroportional to radius) were used the natural tendency for separation to occur on the outer walli, - as exhibited by irrctational flow, would be over- come when the absolute velocity vectors at the outside diamster of the inlet were between 15 and 20 deg off the axis. Accordingly, a set of inlet guide vanes was designed for use with the swirl- “type header which would set up a 20-deg solid-body rotation. The guide vanes were designed so that the equivalent cone angle of the intervening pas- sages would be 10 deg, with the schedule of di- vergence following the relation dP/dx = ag clong a 2/1 ellipitical mean line, The resulting blades had blunt trailing edges that blocked approximately 17% of the inlet passage area. The trailing edge area was distributed so that the blockage occurred in the mid-passage region, with no blockage next to the walls. As was expected, the inlet guide vane system gave flow reversal along the inner wall, This reversal was eliminated by o drag ring which biocked part of the inlet area af the blade trailing The size ond location of the ring were This combination of edges. determined by experiment. header, core shape, and inlet guide vane system gave flow in which the throughput component was not stagnated or reversed of any point on either wall. The pressure loss across the inlet guide vane system was less than that obtained with no guide vanes. In other words, the inlet guide vanes recovered part of the inlet velocity head. The re- fations between the inlet headers, inlet guide vanes, and core are shown in Fig. 13. ‘ The flow problem is made particularly difficult by the design having to be evaluated by experimental tests. Vorious techniques, including pitot trav- "erses, flow visuvalization through the use of dye injections, and conductivity probe measurements on sait injections, are being used. Tests that are felt to be definitive are being carried out on the most promising designs, both with and without vanes, in a one-half-scale model. The volume heat source of the reactor core is simulated with electrically heated sulfuric acid. While the results of the tesis made fo date are still being studied, and they certainly pcse ques- tions that have yet to be resclved, it does appear that the system performs better without the inlet guide vanes than with them. The principal problem either with or without the vanes appears to be that of temperature fluctuations at the wall caused by eddying of the fuel., An experimental evaluation of the amplitude that can be tolerated for such fluc- tuations is under way, i CRNL—LR—DWG 15014 Fig. 13. Diagram of Core Flow System Showing Relations Between the Inlet Headers, Inlet Guide Vanes, and Core. Pumps and Expansion Taonk Various types of pumps were considered for use in the high-temperature-liquid systems. Those seri- ously considered included conventional centrifugal and oaxial-flow pumps and electromagnetic and canned-rotor pumps. Many of these pumps were tested ot ORNL, and some performed quite suc- cessfully.” Each type was found to have advan- tages and disadvantages that had to be evaluated 7E, s. Farris, Summary of High Temperature, Liquid Metal, Fused Salt Pump Development Work in the ORNL-ANP Project for the Period [July 1950-jan, 1954, ORNL CF-54-8-234 (Auvg. 1954). 29 according fo the operating requirements of high- temperature-liquid systems for aircraft instolilations. A most important requirement of an aircraft type of pump is that its weight be reasonable. From the standpoint of the aircraft designer the weight should include all the equipment required to drive the pump; that is, if an electric motor were used, the weight of the electrical generctor should also be included. The importance of the weight of the drive equipment makes the efficiency of the pump alsc an important consideration. The combination of these two factors eliminates electromagnetic pumps from consideration for aircraft applications, because the weight of an electromagnetic pump is inherently from 100 tc 1000 times greater than that of a centrifugal pump driven by a hydraulic motor or an air turbine. it appecrs that a centrifugal-pump drive-system weight of about 1 lb/hp can be ob- tained with an air bleed-coff turbine type of system. A further disadvantage of electromagnetic pumps is thot they are not suitable for operation with molten salts. Canned-rotor pumps were also considered, but they have the same disadvantage thet the electro- magnetic pump has of requiring heavy electrical generators. A further disadvantage is that the thin sheil required between the rotor and the field magnets constitutes a frangible diaphragm in the system. A very large jet of fluid may be ejected when such @ rupture occurs, and a serious fire would result. Pumps that do not require frangible diaphragms in the system may give trouble by producing small lecks which would be annoying and troublesome, but such difficulties are relatively trivial when compared with major abrupt ruptures. Even though the canned-rotor type of pump is suit- able tor operation with molten salts ond even though failure of a canned-rotor pump diaphragm should not lead to a sericus fire with molten salts, the large amounts of radicactivity that would be expected in the molten salts in a full-scale reactor would constitute o for more serious hazard than the fire associated with sodium or NeK. Thus it ap- pears that neither the electromagnetic nor the canned-rofor type of pump is well suited to nuclear aircraft application. Quite a variety of mechanical pumps was con- sidered, but the mixed-flow and radial-flow types of centrifugel pump seem to be much the best adapted to aircraft requirements. Since it is es- sential that the reactor core and the heat ex- changers be as compact as possible, it is neces- 30 sary to mcake use of rather high pressure drops through these components. This in turn means that the pump heads must be between 30 and 300 ft. Therefore if an axial-flow type of pump were em- ployed, it would be necessary to use multiple stages, Flow rates of 500 to 5000 gpm will be required, and thus relatively high shoft speeds are essential if the impeller diameter is to be kept 1o ¢ to 12 in. [mpellers of this small size are es- sential if the installation is to be kept reasoncbly cempact. ' The principal problem in the design and develop- ment of a centrifugal pump for high-temperature liquids is the shaft seal, and quite a number of seals were considered.?® One of the first con- sidered was a graphite-asbestos packing placed around the pump shait, with the gland either in the fluid being pumped or in the gas space chove the fluid in a sump pump. This type of seal tends to give a considerably higher leakage rate than is acceptable and a relatively short shoft life, since the shaft wear is substantial, It works tolerably well when used above the gas space in the sump pump; but, if the seal is placed in the pumped fluid so that there is seepage through it, oxidation of the fluid takes place ot the cutboard end of the seal, and high shaft wear and corrosion rates are likely fo result, An unusual type of seal that has received o con- sidercble amount of attention is the frozen seal.” This type of seal was first developed for use with sedium, |t depends on the use of a cooled gland arcund the shaft that is flocded with the fluid being pumped, Friction between the pump shaft and the frozen fluid in the gland is sufficient to melt o very thin film at the shaft surface. A seal of this type works well with sodium, because the shear strength of the sodium is only of the order of 50 to 100 psi at room temperature, and aiso well with fead, which has a shecr strength of several hundred pounds per square inch at a temperature of arcund 200°F. The freezing temperatures of sodium and lead can be readily oktained with a water-cooled glond, Efforts to make this type of seal work with Nak hove been unsuccessful because of the seal gland having to be cooled to well below the eu- tetic temperature (about —15°F). Unfortunately, the hardness values of the fluoride fuels are muych higher in the temperoture range immediately below 8. E. Schmitz, Trans. Am. Soc. Mech, Engrs. 71, 635 (1949). their melting points thon the hordness value for sodivm. Frozen seal experiments have been made with many fluoride mixtures, including many glassy melfs with large percentages of beryllium fluoride, but in all instances serious cutting of the shaft occurred within a few hundred hours. In addition, large amounts of power were required for contin- vous cperation, and very lorge amounts of torque were necessary in order to breck the frozen seal during startup of the pump; pumps normally re- guiring only 3- to 5-hp motors were found to require 30- to 50-hp motors to be started. The face type of seal used widely in automobile water pumps, refrigerant pumps, end domestic water pumps appears to be a promising solution to the seal problem. [n most applications it is flooded with the pumped fluid during operation. Its very low leakage rate and long fife depend on the mating faces of the seal being finished to essentially eptically flat surfaces. The liguid tends to fill the space befween the two seal surfaces ond to form a meniscus between the edges of the seal faces on the gas side. The surface tension in the meniscus across this very narrow gap gives a pressure in the fluid between the seal faces that is sufficient to hold the surfaces aparf, Therefore the surfaces do not contact each other, but, rather, they shear the fluid film between them. Thus the seal surfaces operate under ideal |ubrication conditions. Seals of this type have opercted for years with no meas- urcbfe wear. They are relatively insensitive to starts and stops if the seal-face pressure is kept fow, It is important that they be adequately cocled and that the product of the pressure in pounds per square inch on the seal face ond the rubbing ve- locity in feet per minute not be excessive. Values ~as high as 100,000 for this pressure-velocity factor have been reported, with the parts giving very satisfactory service life, [t is essential that the mating surfaces in a seal of this fype be compatible from the standpoint of boundary lubrication, or else they moy be scored during storts and stops, It is also very impertant that this type of seal be mounted on a shaft that runs true, with a minimum of vibration. This means that the shaft must be well balanced with the impeller and bearing as- sembly, that the radial lcoseness in the bearings must be small, and that the seal must be mounted in such @ way that it is both concentric and square with the axis of the shaft, Also essential is flexibility in the seal mounting, which can be readily obtained either with a corrugated diaphragm or a bellows, the latter being the more commonly used, Since graphite is a porous material, the graphite washers used in seals of this type are ordinarily impregnated with materials suck as a plastic, lead, babbit, silver, ete. For high-temper- ature use in the ART fuel pumps the seal will operate with inert gas on one side and a flood of oil on the other, rather than fuel, to prevert con- tomination of the fuel. The bearing problem has much in common with the seal problem for high-temperature-fluid pumps, In all instances the fluids are very corrosive to most materials, and therefore only a few materials, such as graphite, certain iron-chrominum-nickel al- lovs, and cemented carbides, can be used. Further, the fluids pumped will remove any adsorbed films such as sulfides, phosphides, etc., that would tend to alleviate boundary-layer [ubrication con- ditions, The viscosity of sodium is about one- fiftieth that of water, while the viscosity of the fluoride fuel mixture is about the same as that of water. Thus neither of these fluids serves as o really good lubricant. While they have the advan- tage of wetting the surfaces of iron-chromium-nickel alloys very effectively, they tend to strip off the protective films that are formed under ordinary con- ditions in most types of petroleum- or other nydro- carbon-lubricated bearings. It is evident that bearings designed to operate in molten metal or molten salt must be lightly loaded and carefully aligned, . The investigations have indicated that the bear- ing and seal should be placed in @ cool zone above the pump impeller with a heat dam between them and the fluid being pumped. This arrangement makes it possible to use conventional bearings and seals. Pumps of this type have proved te be quite satisfactory and have the advantage of being relatively insensitive to the type of fluid being pumped; that is, the same type of pump can be used for sodium, NaK, the fivoride fuel mixture, lead, sodium hydroxide, or other fluids, Two fuel pumps and two sodium pumps will be located at the top of the reactor, These pumps will be similar, but the fuel pumps will have the larger flow capacity. The fuel pumps will also include xenon-removal systems in which most of the xenon and krypton and prebably some of the other gaseous fission-preduct poisons will be removed from the fluoride fuel mixture by scrubbing with helium, The fuel will enter the xenon-removal system from the eye of the pump and will pass up the center of the 31 shoft and inte the mixing chamber. In the mixing chamber the fuel wiil spray through a helium at- mosphere and impinge on the wall of the chamber, The resulting mixture will be very foamy and wili have a latge gas interface. The helium will enter the system just below the shaft seal and flow down the annulus around the shaft, through the upper slinger vanes, and into the mixing chamber, The fuel-helium mixture will then be pumped into the fuel expansion volume, where the helium centaining the xenon, krypton, and other fission-product gases will be removed via the off-gas system. The circuit wiil be completed when the fuel leaves the expan- sion volume by grovity flow intc the centrifuge, where any entrained gas will be removed and then be returned to the expansion volume. The fuel will be pumped through the centrifuge heles and will re-enter the main fuel system on the discharge side of the fuel pump. The system is illustrated in Fig. 14. Experimental results obtained with a single pump operating in a test lcop showed that the fuel level in the expansion chamber and the rate of bleed flow affect the ability of the centrifuge to prevent helium bubbles from entering the main fuel system. The system operates very well with a fuel level of ¥ in. in the expansion chamber and a bleed flow of up to 13 gpm. With a fuel level of 3 in. in the expansion chamber the system can be operated with a bleed flow of 25 gpm. The design value for the bypass tlow rate is 12 gpm. Similar tests on g full-scale aluminum model of the fuel pump—expansion tank region have given similar results, A number of other special features have been in- cluded in the pump design for adaptation to the full-scale reactor shield. The pump has been de- signed so that it can be removed or instailed as a subassembly, with the impelier, shaft, seal, and bearing comprising a single compact unit. This assembly will fit into a cylindrical casing welded to the top of the reactor pressure shell. A 3-in. tayer of gemma-ray shielding just above a l/z-in., layer of zirconium oxide around the lower part of the impeller shaft will be at the same level as the reactor gamma-ray shield just outside the pressure shell. The space between the bearings will be filled with oil to avoid a gap in the neutron shield. The pumps will be powered by hydroulic drive units in order to provide good speed centrol, along with compact, reliable metors. 32 Fill-and-Drain System {Including Enricher) The system designed for filling the reactor with the barren fuel-carrier sait NaZrF, for adding the enriched uranium-bearing fuel component Na,UF , and for draining the reacter is described schemat- ically in Fig. 15. For the initial fiiling operation the fuel carrier will be added to the fill-and-drain tank at a temperature of about 1200°F and will be pressurized into the reactor @ number of times at this temperature. The reactor cond the fill-and- drain tank will have been preheated by cperation of their respective NaK systems. Approximately 60% of the uranium-bearing component of the fuel will then be added from a simple melt pet by gas pressure transfer. The remaining uranium-bearing material required to achieve criticality will be added in small increments by using the enricher device shown in Fig. 15. The fuel fill-and-drain system is designed to permit the transfer of fuei to and from the reactor after the reactor system has been completely assembled. A heating system is included to maintain the fuel at a temperature above its melting point. This same system aiso serves to cool the fuel by removing decay heat from fuel drained after the reactor has been operated at power. The initial fuel charge is to be admitted to the fuel fili-and-drain tank through a fill nozzle provided at the top of the tank. This initial filling operation will be carried out by persennel who will be inside the pressure vessel ot the fill-and-drain tank. After the addition of the barren salt to the {ill-and-drain tank, the system will deliver this salt into the reactor for initial testing. The barren saft will then be drained for the addition of the enriching mixture. The system will then be used to fill or partly fill and drain the reactor ¢ number of times for mixing the salt and enriching mixture. This enriching and mixing cycle may have to be repeated a great many times. Fuel flow into the reactor will be by displacement with helium, and flow back to the fill-and-drain tank will be by gravity. The helium displaced from the reactor duting fuel addition will be discharged through the off-gas system. Helivm displaced from the drain tank as the fuel is drained will flow through part of the off-gas piping intc the top of the reacter, The fili-and-drain tank will also receive the fuel charge from the reacter ot any time during SECHET ORNL~LR-DWG 16015 EXPANSION VOLUME H— SLINGER IMPELLER P ] EXPANSICN VOLUME [—MIXING CHAMBIZR _}—SEAL PLATE — ] ]///—_ CENTRIFUGE SEAL VANE BAFFLE ——} —— CENTRIFUGE HOLZ PUMP N\ piscHaRrGe | FUEL PUMP X/ / / PUMP SUCTION | | Fig. 14. Section Through Fuel Pump~Expansion Tank Region Showing Xenon-Removal System of the Fuel Pump. 33 SECTPT ORNL-LR-[CWG {6046 HELIUM ____,[ J__, OVERFLOW LINE i Ps ¥ - VAPOR b ENRICHER T BREAK POINTS V" - ey HE..IUVI TO POSTPOWER o FILL - AND - DRAIN SAMPLER R PRtP(,Wr_R TANK TO FUEL PROCESSING SAMELER SYSTEM TANK " TANK FOR ADDITION OF CARRIER AND INITIAL BATGH OF CONGENTRATE Fig. 15. Schematic Diegrom of Fuel Fill-and- Drain System, Including Enricher. the operation when an emergency requires that the fuel be drained. Under these conditions, heat will have to be removed from the fuel ot o rate equal to the fission-product-decay heat-liberation rate. Finally, the fuel fill-and-drain system will receive the fuel from the reactor at the completion of the test and will hold it as long as decay heat requires cooling of the fuel. It will then be dis- placed by helium pressure into the fuel recovery tank for removal. No provision has been made for transferring the fuel back frem the recovery tank to the fill-and-drcin tank or for cocling the fuel in the recovery tank. After the preliminary design studies were com- pleted, it was decided that the fill-and-drain tank should be designed with two independent sets of cooling tubes, each designed to permeste com- pletely the fuel veclume. This arrahgement will permit the tank to be cooled even though some component of one of the two cooling systems, such as a pump, radiator, or pipe, should fail. The 34 design heat load on the tank is 1.75 Mw. The theoretical decay heat ot the instant of shutdown in a 60-Mw reactor is given as about 3.6 Mw, but only a 1.75-Mw cocling system is required, because the fuel will not be drained into the drain tank until the fission heat is negligible and untii the fuel has been cocled to below 1200°F. The drain valves will not be opened until at least 8 sec after the contre! rod has been fully inserted. The design is based upon the assumption that such emergency fuel drainage would ecccur after contin- vous operation at 60 Mw for one month. The heat capacities of the fuel and mechanical equipment are great enough tc absorb the decay heat above 1.75 Mw for the few minutes during which the decay heat generation would exceed this rate, The fuel piping between the reactor and the fili-and-drain tank is designed to drain all the fuel from the reactor under gravity flow conditions in not more than 3 min. [t is calculated that this condition will be met even though one of the two drain valves cannot be opened. The requirement for a completely reliable cocling system seems to be met by the use of bundles of cooling tubes inserted into the tank. In order to provide cooling for all metal surfaces in contact with the fuel, the tank and the standpipe to the reactor are both jacketed. The coolant to be used for this service will be NaK (56-44 wt %). To provide two independent sets of cooling passages, each of which permeates the entire fuel volume, a cylindrical tank was selected, with each end serving as a tube sheet. Horizontai decks of U-tubes arranged on each tube sheet in a square-hcle pattern provide alternate layers that originate at opposite ends of the rank (Fig. 16}. Design calculations indicate that fellowing an emergency drainage of the fuel, the cooling system will, even with only one circuit functioning, limit the fuel temperature in the tank to 1800°F. When both coolant circuits function, the maximum temperature at any point in the fill-and-drain tank will be limited to 1600°F. The jackets around the cylindrical shell of the tank and around the drain line, however, are not served by two separate systems. Coolant flow through the tank jocket wiil be in parallet with coolant flow in cne set of U-tubes., The NaK will enter the jocket around one half of the tank circumference and will return to the coolant channe! cround the other half of the tube sheet periphery. Flow guides are placed along the horizontal, CEONFBENRAL ORNL—~LR-~DWG 16!26 Nak SUPPLY -1 j, {HHOG°F) ¥ FUEL DRAIN ¢ ° LINE . “~--INNER CYLINGER {He BLANKET AT 5 psi } -~ BAFFLES { FOR SUPPORT - NaK SUPPLY —2 AND FLOW DIVISION) L} SUPPORT CYLINDER 1 (5000 1b UPLOAD} Fig. 16. Fuel System. Fill-end-Drein Tank Cooling annular jacket to distribute the coclant flow as required. The two sets of U-tubes ore arranged so that one set cools the fuel zone just inside the cylin- drical tank wall more thoroughly than the clternate set, and the coolant supply to the jacket is from the alternate system, which is least effective in cooling the tank wall. At one end of the tank, where the fuel shell and tube sheet intersect, a triangular “filler ring is placed to displace fuel from this corner zone, where coocling would be poor if one of the two coolant circuits failed, At this end of the tank, failure of the coolant circuit would interrupt cooling behind the tube sheet and in the jacket. At the other end of the tank, coeling would continue either in the jacket or behind the tube sheet, and no filler ring is required. Further details of the operation of this system are given in the section “Operation of the ART" and on the flowsheets in Appendix A. Sampling System Samples of the barren molten salt will be token immediately aofter ‘‘shakedown’’ operation of the ART by pressurization of the melten salt through a heated Inconel tube from the fill-and-drain tank into a detachable sample receiver. Additional somples will be taken during enrichment in the same manner fo defermine the homogeneity of mixing and the concentration of uranium in the mixture, Samples of the molten fuel mixture will also be taken after nuclear power operation. These samples will aiso be transferred through hected Inconel lines into sample receivers, but, because of the high level of radicactivity in the fuel, the operation is to be accomplished by remote control, and the lines and receivers must be shielded. To prevent premature activation of the sampling valve, a frangible disk valve will be included in each sample transfer line. A total of five postpower samples may be taken. The five receivers wiil be encased in the same lead shielding urit, which will be removed after the entire ART opercation is completed. Fuel Recovery System After completion of ART operation, the fuel will be drained into the fill-and-drain tank and held there for a period of several days to allow fission- product decay and the consequent lowering of decay heat production. The fuel will then be transferred into the fuel recovery tank for delivery to the recovery and reprocessing facility. The fuel recovery tank will consist of four, interconnected, approximately 9-ft lengths of 8-in.- IPS Inconel pipe furnished with elecirical heaters. The assembly is designed for cooling by radiation ond natural convection to the atmosphere. It will be shielded with approximately 10 in. of lead. The molten fuel will be transferred to the tank through a heated Inconel line containing a frengible disk valve and an open bismuth valve. When the transfer is completed, the bismuth valve will be closed and both the transfer line and the helium line will be severed, as will all heater and thermo- couple leads, The assembly will then be lifted out of the cell and loaded on a truck for delivery to the recovery facility. The assembly will include o transfer line and the other necessary service connections for use in the course of recovery operations. HEAT EXCHANGERS AND HEAT DUMPS The spherical-shell fuel-to-NoK heat exchonger, which mokes possible the compact layout of the reactor heat exchanger assembly, is based on the 35 use of tube bundies curved in such a way that the tube spacing will be uniform, irrespective of lotitude.® The individual tube bundles terminate in header drums. This orrangement focilitates assembly because small tube-to-header bundles can be assembled, made leaktight, and inspected much more easily than can one large unit. Fur- thermore, these tube bundles give a rugged, flexible construction that is admirably suited to service in which large omounts of differential thermal expansion must be expected, The config- uration of the heat exchanger tube bundles is iHustrated in Fig. 17. The heat removed from the fuel by the NaK wili be transferred to air in the NaK-to-agir radiators, which will be installed in ¢ hect-dump system designed to simulate the turbojet engines of the full-scale aircraft in a number of important respects, such as thermal inertia, NaX holdup, and basic fabricational methods. The round-tube and plate- fin radiator cores are fabricated of type 310 stainless-steel-clad copper fins spaced 15 per inch and mounted on %6-En.=OD tnconel tubes placed en 2/3=-irx., square cenfers. 1 he tubes are welded ond brazed into round header drums. The individual radictor cores consist of twe halves 15 by 30 in. The fin matrix depth in the air flow direction is 5.33 in. A typical radiator is illustrated in Fig. 18. The basic reguirement of the heat-dump system is to provide heat-dump capacity equivalent to 40 Mw of heat with ¢ mecn temperature level of 1300°F in the NaK system. The NaK will be circuiated through eight separate systems. Four will constitute the main heat-dump system, two witl serve the reflector-moderator cooling system, and two will serve the fuel fitl-and-drain tank. in the main system a group of four NaK fill-and- drain tanks will be used. The tanks will be pres- surized to force the NaK into the main cooling circuit. The 12 tube bundles of the fuel-to-NaK heat exchanger will be manifolded in four groups of three each. The NaK will flow from these tube bundles out to the radiators, which will be arranged in four vertical banks with four radiotors in each bank. The NaK will flow upward through ihe radiator bank to the pumps. A small NaK bypass fiow around the radiators will pass threugh o cold trap in order to maintain o low oxygen concen- %A. P. Frags and M. E. LaVerne, feat Exchanger Design Charts, ORNL-133C (Dec. 7, 1952). 36 tration in the NaK. (For details of the NaK systems see Appendix A, Flow Diagram 6.} The two independent reflector-moderator heat- dump systems (referred tc on Flow Diagram 6, Appendix A, as the Auxiliary System) will be essentially similar to the main NaK system except that their combined capacity will be about one- third thot of one of the four circuits of the main heat-dump system. The NaK will be circulated to both the Na-to-NaK heat exchengers in the top of the reactor, where the two NaK streems will pick up from the sodium the heat generated in the island and the reflector. The twe NaK streams will pass to small NaK-to-gir radiators, where they will be cooled ond returned to their respective pump suctions. A bypass cold trap will be inciuded in each system, as in the main NoK systems, while a single fill-and-drain tank will serve both systems. Four axial-flow blowers will force 300,000 cfm of air {which expands to 6.7 x 10° cfm when heated to 750°F) through the radicfors and out through a 10-fi-dia dischorge stack 78 ft high. Since the axial-flow blowers will stafl and surge if throttled, contrel will be accomplished threcugh bypassing @ portion of the air around the radiators. The heat- dump rate will be modulated by varying the amount of air bypassed through a set of conirollable jouvers mounted in such & way as to bleed air from the plenum chamber between the blowers ond the radiators. The arrangement will include louvers te block off the air passage to the radiators and louvers in the bypass duct; one set will be opening while the other is closing. This arrange- ment should give good centrol of the heat load from zero to 110% of the design load. Since each biower will be driven with an a-¢ motor independ- ently of the others, the heat-dump capacity can also be increased in increments of 25% from zero to full load by changing the number of blowers used. Heat barriers mounied on both sides of the ra- diators will be required in order to minimize heat losses during wormup operatiens. Warmup will be” accomplished by energizing the NaK pumps and driving them ot part or full speed. As a result of fluid frictional losses approximately 400 hp must he put into the pumps in the NaK circuits and will appear as heat in the fluid pumped. A mechanical power input of 400 hp to the NaK pumps will sroduce a heat input in the Nall system of approxi- mately 300 kw. This should be encugh to heat the system quite satisfacterily if the radiator O o @ o O bed O - O I ' p 4 o o Yo R PV, MNaK Heot Exchanger Channel. af Ge in Fuel Mode!l of Ma . 7. g @ F 37 | UNCLASSIFIED 1 Fig. 18. Prototype ART NaK.tc-Air Radiator. cores are blanketed to prevent excessive heat losses. However, electrical heaters will be avail- able on the NaK lines so that the NaK pumps can be run at one-half speed to reduce stresses and weor during zero- ond low-power operation. Rela- tively simple sheet stainless steel doors with 3.0 in. of thermal insulation will, when closed over both faces of a radiator (100-ft2 inlet-face area) filled with 1100°F NaK, reduce the heat loss there to about 30 kw. The heat appearing in the reflector-moderator will be cbout 3.5% of the reactor power oufput. The cooling circuit will also remove heat from the core shells ond the pressure shell, and therefore the total amount of heat to be removed from this cooling circuit will be ebout 10% of the reactor output. This must be removed at a mean NaK circuit temperature of about 1050°F. Four ra- diators that have inlet faces 2 x 2.5 ft each and the same basic geometry as that ysed for the main heat dumps wiil be employed. These radiators will be equipped with louvers and heat-barrier doors like those used on the main NaK radiators and will be supplied with air from the same air duct {see Flow Diagram 6, Appendix A). 38 The two independent heat-dump systems for heating and cooling the fuel fill-and-drain tank (referred to on Flow Diagram 6, Appendix A, as the Special Systems) will also be essentially similar to the main NaK system. The two NaK-to- air radiators in these systems will be served by individual air blowers, When the systems are to be used for adding heat to the fuel in the fill-and- drain tank to preheat the tank for filling or to maintain the fuel above its melting temperature, they will be capable of heating the tank to 1250°F. The systems will also be ready at any time to remove fission-product-decay heat from the fuel. OFF-GAS SYSTEM The fission-product gases which will be purged from the fuel in the fuel expaonsion tank are to be centinvously removed from the fuel system to a water-cooled charcoal adsorber, in which the gases will, in effect, be “‘held up’’ « sufficiently fong period of time for safe discharge to the stack. In the fuel expansion tank the gases will be diluted with 1000 to 5000 liters of helium per day and will be bled first through a water-cooled vapor trap (te remove ZrF, vapor) and then outside the reactor and cell through an empty pipe into a charcoal- filled pipe in a water-filled tank. The empty pipe was provided to permit some decay of the gaseous activity so that when it is subsequently adsorbed by the charcoal the resuitant heat can be satisfac- torily transferred to the water surrounding the charcoal-filled pipe. There will be sufficient charcoal to assure a 48-hr holdup of K¢88, which will be the only gaseous fission product with significant activity after passage through the charcoal. Experimental work has indicated that 64 13 of charcoal will be required. Two parallel charcoal adsorber systems are provided. Provisions have also been made for bleeding the cell atmosphere through an auxiliary charcoal bed in the event that a leak occurs in the reactor off-gas system within the cell. The charcoal bed in this auxiliary vent system will be bypassed during normal operation for disposing of insfrument bleed gases. Four bypass loops will be provided in the two off-gas systems, one before and one after the charcoal section of each system, so that the activity in the isolated gas samples can be counted. Further, the vent lines to the stack will be mon- itored to determine whether the gas may be safely discharged. The system will operate at o pres- sure slightly above atmospheric and will not require pumps. Except for the equipment within the reacter cell and in the water tank (containing the empty and the charcoal-filled pipes), which will be buried in the ground, all components of the off-gas system are to be housed in a special building provided for that purpose at the southwest corner of the ART building. A fundamental design criterion is that access to this “‘off-gas shack’ should not be limited by radicactivity from any other part of the system but that the shack should be isclated in the event of a leck of gaseous activity therein. In the design of the off-gas system the calcu- lations that were made!®~12 were strictly theoret- ical, but they were strengthened by the operating experience which had been obicined on a similar adsorber system used with the HRE. In particular, flow vs pressure-drop data and flow vs breakthreugh times in charcoal adsorbers were obtained from HRE experience.'S FExperimental data'4 on the adsorptivity of charcoal indicate that the design of the ART adsorber is conservative. Details of the system are presented on Flow Diegrem 3 in Appendix A. AIRCRAFT-TYPE SHIELD The mest promising of the several shielding arrangements that were considered for the ART seems to be the one which is functionally the same as the arrangement for an aircraft requiring a unit shield — a shield designed to give ~10 r/hr at 50 ft from the center of the reactor. Such a shield is not far from being both the lightest and the most compact that has been devised. It will make use of noncritical moterials that are in good supply, and it will provide useful performance data on the effects on the radiation dose fevels of the release of delayed neutrons and decay gammas in the heat exchanger, the generation of secondary gammas throughout the shield, etc. While the complication Wy, B. Cottrell et i, Aircraft Reactor Test Hazards Summary Report, ORNL.-1835, p 24 (Jan. 19, 1955). ”Ca S. Burtnette, Fission Product Heating in Off-Gas System of the ART, ORNL CF-55-3-191 (March 28, 1955). 12, B. Andersen, Adsorption Holdup of Radicactive Krypton and Xenon, ORNL CF-55-8103 (Aug. 154, 1955). mh Spiewak, Use of HRE Charcoal Adsorbers in the HRT, ORNL CF-54-7-26 (July 8, 1954). Yy, E. Browning and C. C. Bolta, ANP Quar. Prog. Rep. March 10, 1956, ORNL.-2061, p 193. ~ spherical-shell of detailed instrumentation within the shield does not appear to be warranted, it will be extremely worth while to obtain radiaticn-dose-level data at representative points around the periphery of the shield, particulorly in the vicinity of the ducts and of the pump and the expansion tank region. The shield for the reactor will have some chorac- teristics that will be peculior to this particuler reactor configurction. The thick reflector was selected on the basis of shielding considerations. Use of o thick reflector is based on twe major reasons: a reflector about 11 in. thick followed by o layer of boron-beoring material will attenucte the neutron flux to the point that the secondary gamma flux can be reduced to a value about equal to that of the core gamma radiation; it will also reduce the neutron leakage flux from the reflector into the heat exchanger tc about the level of that from the delayed neutrons which will appear in the heat exchanger from the circuleting fuel., An additional advantage of the thick reflector is that 99.8% of the energy developed in the core will appear as heat in the high-temperature zone within the pressure shell. This means that very little of the energy produced by the reactor will have to be disposed of with a perasitic cocling system at a low-temperature level. The material in the intermediate heat exchanger is about 70% as effective as water for the removal of fast neutrons; so it too is of value from the shielding standpcint. The delayed neutrors from the circulating fue! in the heat exchonger region might appear to pose a serious handicap. However, they will have an attenuation length much shorter than the corresponding attenuation length for radiation from the core. Thus, from the ocuter surface of the shield, the intermediate heat ex- changer will appear as a much less intense source of neutrons than the more deeply buried reactor core. The fission-product-decay gammas from the heat exchanger will be abeut equolly as important as the secondary gammas from the beryllium and the reflector shell. Thermal insvlation 0.5 in. thick will separate the hot reactor pressure shell from the gamma shield, which will be o loyer of lead 4.3 in. thick. The lead, in turn, will be surrounded by a 32-in.- thick region of borated water. The slightly pres- surized water shield will be contgined in an aluminum tonk. Cooling of the lead shield will be effecied by circulation of woter through coils embedded in the lead. The borated water shield 39 will also be ccoled by water circulated in coils submerged in the borated water and by thermal convection of the atmesphere in the reactor as- sembly cell. An opening will be made at the bottom of the shield for filling and draining of the reactor. The fuel fill-and-drain tank will be shielded with 10 in. of lead to reduce the dose to 1 r/hr at the shield surface one week after full-power operation. The resulting shield weight for a 13-f° capacity tank will be about 30 tons. CONTROLS AND INSTRUMENTATION The early effort by ORNL personnel to develop the circulating-fuel type of aircroft reactor was motivated in part by o desirable control feature of such reactors — the inherent stability of the reacfor ot design point that results from the neg- ative fuel and over-all temperature coefficients of reactivity. in a power plant with this characteristic the nuclear power source will be a slove to the turbojet load with but a minimum of external control devices. This predicted master-slave relation between the foad and the power source was verified by the ARE. Work with an electronic simulator indicaftes that the ART should behave in essentially the same way. Controlwise, the power plant consists of the nuclear source, the heat dump (in the case of the ART), and the coupiing between source and sink (the NaK circuit). Control at design point can be effected to some extent by nuclear means at the reactor, by changing the coupling (i.e., changing the NaK flow) or by changing the load (i.e., the heat dump from the NaK radiators). For the ART ot design point the regulating rod wifl be used mainly for adjusting the reactor mean fuel temperature., In particular, an upper tempera- ture limit will cause the regulating rod to insert, and therefore the fuel outlet temperature should not appreciably exceed 1600°F, even in transients. This limit will override any normal demand for rod withdrawal. Furthermore o low NagK ocutlet tempercture from the heat-dump radicters will avtomatically decrease the heat lcad to keep the lowest NaK temperature of the system ot no less than 1070°F. This lower temperature limit will override all other demands for power, Control of the ART falls in thiee different categories of operation: stertup, eoperation be- tween startup and appreciable power (about 10% of design point), and operation in the range from 40 10% to full power. For the second and third of these categories the nature of the reactor and power plant is so different from that of conven- tional high-flux reactors that control must be based on inherent characteristics of the reactor to a large extent rather than on conventional reactor-control practice. Control at startup utilizes, in principle, old reactor-control practice with short-period ‘“’scrams’’ that are conventional, Experimentation will take place primarily in the startup and design-point regions. In the inter- mediate region between these two, little testing will be done. Fission chambers and compensated ion chambers witl be locoted beneath the reactor shell just outside the lead region. The region around the fill-and-drain pipe will be fililed with moderator material through which cylindrical holes for these chambers will run radially out on lines which intersect at a point below the center of the reactor, The chamber sensitivities will be adequate for the entire range of nuclear operations. The contrel system is designed to provide auto- matic corrective action for emergencies requiring action too rapid to permit operator deliberation. Automatic interfocks will prevent inadvertent dangercus operation, with minimum operator limita- tion. Operation in the design power range will be independent of nuclear instrumentation during power transients. Three classes of emergencies have been provided for. Class I. — Any of the following events will cause the rod to be automatically and completely inserted at a rate of 1% Ak/k in 8 sec: I. stoppage of either sedium pump, 2. stoppage of either fuel pump, 3. drop in fuel or sodium liquid levels in the ex- pansion tanks when pumps are operating af design peint, 4. failure of the oil system to the sodium or fuel pumps, 5. leakage of fuel into NaK or NaK into fuel, 6. failure of commercial power or locally gen- erated power. At the same time one-half the blowers will be shut off to reduce the heat load. The load will then be further reduced by the radiator shutter auto- matically closing in response to a 1070°F low- signal for the NoK-te-air radiator outlet Dumping of the fuel will not be automatic but will probably be initicted by the operdator. limit temperature. Class I, — Any of the following events will cause the rod to be automatically inserted ot o rate of 1% Ak/k in 8 sec: 1. stoppage of any NaK cireuit, 2. leakage of NaK to the atmosphere, 3. trouble in the off-gas system (to be defined), 4. maximum fuel temperature greater than 1650°F, 5. maximum 1300°F, 6. maximum sodium temperature greater than 1250°F when either set of sodium-to-NaK radiator lcuvers is wide open, 7. reactor on positive period of less than 3 sec, 8. fill-and-drain tank mecan temperature grecter than 1300°F or less than 1100°F, 9. failure of the oil system to any NaK gump, 10. rod-drive trouble. Class I, —~ |t is planned that the operator will be warned of any of the following events, but no automatic corrective action will take place: sodium temperature greater than 1. leakage of sodium into fuel or fuel into sedium, 2. lowering of the water level in the outer cell, 3. excessive radiation level in the cell as deter- mined by monitors, excessive humidity in the cell, excessive rod temperature, excessive chamber temperature, . oxygen in the cell. Lists of the instruments for the ART have been prepared that give the foliowing information for each required measurement: type of instrument pickup, location of pickup, type of presentation, reading range, and accuracy. |he necessify for using an instrument was determined by whether it would be required for the safe and orderly conduct of the test, for providing sufficient information for evaluation of test results, and for previding not otherwise available, Five stations have been provided in the ART building at which instruments will be read: the control room, the information room, the auxiliary equipment panel, the vent house, and the temporary panel for fuel sampling and recovery. ~ All instruments pertinent to the nuclear perform- ance of the reactor, as well as to the control of equipment which coffects nuclear performance, are located in the control room. The instruments required for determination of reactor power and heat exchanger and radiator performance are located in the information room. In general, operating instruments for auxiliary systems, such as water, hydraulic fluid, lubricating SO A location, informetion all process oil, gas, etc., as well as for the NaK system and all pumps, will be confined to the auxiliery equip- ment panel. A few pertinent instruments, as well as a number of alarms, are duplicated in both the control room and the cuxiliary equipment panel. With a few minor exceptions, all heater controls and associated femperature instruments are located in the basement on the cuxiliary equipment panel. Off-gos system temperatures will be recorded in the vent house to avcid lines being run to the information room. The temporary panel (for fuel sempling ond recovery) will be located on the main floor just outside the cell and will be connected by temporary lines to equipment within the cell as required to effect the fuel sampling and fuel recovery operations. AUXILIARY SYSTEMS Heliom Supply System Helium is required in the ART principaily be- cause of a need for a flushing gas for removing fission gases, for an inert atmosphere over the tuet, medium for forcing fuel and NaK to and from their respective fill-and-drain tanks inte the recctor system. Since no two of these three uses of helium will be concurrent, the helium consumption rate will be tow. Therefore the helium is to be supplied from 12 cylinders manifolded into two banks by a high-pressure manifeld with isolating valves and pressure-reducing valves for each bank. This will allow use from either of the two banks while the depleted bank is being reploced. (For details of this system see Flow Diagram 9, Appendix A.) sodivm, and NaK, and for o pressurizing Nitrogen Supply System Nitrogen will be required for filling the recctor cell and for operating the pneumatic instruments. The nitrogen atmosphere in the cell is a safety measure. In the event of a simultaneous sodium and water leak from the reactor into the cell, it is important thot the oxygen concentration of the ambient atmosphere be kept low encugh for no detonation of the hydrogen-oxygen mixture to occur. fhe lower combustibility limit for hycrogen- nitrogen-oxygen mixtures is obtained at 5% oxygen. To ensure a substontial margin of safety, it has been decided that the oxygen concentration in the reactor cell will be kept to less than 1%. The oxygen will be removed ‘initially from the cefl by the cell pressure being reduced with o 41 vacuum pump to a maximum of 0.74 psia and the cell being repressurized with dry nitrogen fto atmospheric pressure. The dry nitrogen bleed from the preumatic instruments will act as a continucus purge, which will keep the oxygen content of the nitrogen low. The bleed flow is estimated to be a maximum of 7 scfm. (For details of this system see Fiow Diagram 10, Appendix A.) A permanent storage tank with a capacity of 10,500 scf ot 1000 psi and a pressure-reducing station with two reducing valves that are available at the site will comprise an adequate system for supplying the instruments with the 7 cfm required during the test. A trailer will be used to fill the reactor cell initially. Electrical Power System, Distribution, and Auxiliery Equipment The electrical power for the ART will be sup- plied by two separate scurces. One will be a commercial (TVA) source and the other will be a set of diesel-driven generators. In case of a power failure for any reason, no effort will be made to continue to operate the reactor at power; that is, a power failure or an equipment failure will lead to an orderly shutdown of the reactor. The main power loads are: 1. the two fuel pumps, 2. the two sedium pumps, 3. the two NaK pumps in the reflector-moderator cooling system, 4. the four NaK pumps in the fuel cooling system, . the two Nai pumps in the fill-and-drain tank cooling system, 6. the four radiater blowers in the main duct, 7. the two radiator blowers in the special duct, §. the battery for control and instrument circuits, 9. the pump lube cil systems, 10. the heater and auxiliary loads. The relation between the various pumps is such n that one fuel pump, one sodium pump, one NaK pump in the reflector-moderator cocling system, two NaK pumps in the fuel coeling system, and two radiator blowers should be connected to one power source. 1[hese major pieces of equipment will not have alternate power sources. A station-type battery will be provided, and the circuit will be arranged so that the battery will float on the line at all times to keep a full charge. The battery wiil be used to supply the necessary control and instrument circuits in case of a power outage. Seme emergency lighting will alsc be fed off the bottery. 42 Fuel and Sodium Pump Lubricating and Coocling Oil Systems The lubricating and cooling oil systems for the fuel and sodium pumps are to be sealed so that they will serve as secondary, or backup, containers for the gaseous radioactive products which might leak from the fuel or sodium systems through the primary rotating face seal between the reactor system cover gas and the lubricating oil surrounding the lower seal. The pressure in the lubricating cil system will be a slave to the pressure of the process system through seal-balancing control for holding the differenticl pressure across the lower seal to a minimum and in @ direction to cause leckage of oil into the system rather than gases into the oil. The oil leakage will be trapped and then removed by a separate system. All parts of the system external to the test cell must be capable of withstanding pressures as high as 200 psig for periods of up te 1 hr without failure. Process Water System The ART process water is to be supplied by an ““open,”’ once-throcugh system with two perallel pumps to supply pressure. There will be one supply header and one exit header, and each will penetrate the cell wall. The process circuits will join these headers within the cell to minimize the number of cell penetrations. The flow through each circuit witl be preset before the cell is sealed, and thus there will be no need for remote conirol or remote flow measurement, The use of the open cycle eliminates the need for additional water pumps, heat exchangers, etc., and the consequent neces- sity of canning the pumps to withstand the csll disaster pressure. Check valves, backed up by motor-driven cutoff valves, will be provided in the infet lines through the cell. The check valves, backed up by the cutoff valves, will prevent the escape of the gaseocus activity that would be present in the water lines within the cell if a reactor catastrophe caused the water lines within the cell to rupture. Provisions will be made to assure uninterrupted flow, since water flow to the lead shielding must be maintained during all periods when the reactor and dump tank are at operating temperature so that the lead wiil not be melted. Water will be required within the cell for filling the reactor water shieid and for ceoling the resctor and fuel-dump-tank lead shields, the two instrument pods, and the reactor and fuel-dump-tank vapor sraps. Outside the cell, water cooling will be provided for the lubrication and hydraulic systems, the charcoal adsorber, the penthouse space cooler, and the NalK cold treps and for filling the water annulus of the cell. THE FACILITY The Building The building constructed in 1952 te house the ARE is being modified to provide the space and facilities required for the ART. An addition has been constructed at the south end of the ARE building to effect a 64-ft extension of the eriginal 106-ft-long building, The shielded reactor assembly will be installed in the containing cell that has been provided in the addition for this purpose. Such an arrangement will permit the use of services and facilities that were provided for the original installation., [tems such as the controf room, of- fices, change roems, toilets, storage orea, water supply, power supply, portions of experimental test pits, access roads, security fencing, and security lighting have been incorporated in ART plans. The plan and section drawings of the facility are shown in Figs. 19 and 20. The floor level of the addition is at the ARE basement-tloor grade (ground level at this end of the building), and the cetl for housing the reactor assembly is sunk in the floor. The reactor cell is located in approx- imately the southwest quarter of a 42-fi-wide by 64-ft-long high~bay extensicn and is directly in fine with the ARE experimental bay. The reactor assembly will be positioned so that the top of the shield will be below building floor efevation. The socuth wall of the ARE experimental bay has been removed, and the overhead crane facility has been revised by the installatien of ¢ 30-ton-capacity crane, in addition to the existing 10-ton c¢rane, tc permit use of the experimental pits for instailation of auxiliory equipment and possibly for underwater reactor disassembly work after reactor operation. Also, the truck dosr in the north wall of the building has been enlarged to provide a large entry door to the ART area. Field maintenance and laboratory facilities have been instailed in the area east of the new bay and south of the low bay of the older part of the building. The Reactor Assembly Celf The cell designed for housing the reactor as- sembly is shown in Fig. 21. The cell consists of an inner and an outer tank. The heat-dump equip- ment will be located outside the cell, but nearby. The space between the two tanks is 36 in. and will be filled with water. The inner tank will be sealed so that it can contain the reactor in an inert atmeosphere of nitrogen at 5 psig. The tank has been built to meet ASME code requirements for unfired pressure vessels designed for a 200- psig operating pressure. The outer tank serves merely as a water container, The inner tank will be approximately 24 ft in diameter with o straight section 12 ft long and a hemispherical bottom and top. The outer tark, which is cylindrical, is 30 ft in diometer and about 47.5 ft in height. When the reactor is to be operated at high power, the space between the tanks and above the inner tank will be filled with wcter for carrying off the heat given off by decay gamma activity in the event of an accident so severe as to cause a meltdown of the reactor, About 13 ft of the outer tonk will be above fioor grade. This portion of the tank, as well as the top hemisphere of the inner tank, will not be attached until completion of the reactor instcliation and preliminary shakedown testing. Since the shielding at the reactor and for other radicactive components will be quite effective, it will be possible for o man to enter the inner tank through a marhole for inspection or repair work, |f the reactor has been operated at moderately high power, the fuel will have to be drained. If the repair work requires a relatively feng time, the nitrogen atmosphere will be replaced with air. The unshielded reactor assembly will weigh approximately 11,500 ib, the lead gamma shield will weigh approximately 26,000 tb, and the water in the shield will weigh approximately 38,500 Ib., The shielded reactor assembiy will be mounted in the inner tank on vertical columns with the reactor off center 3 ft from the vessel axis and abcut 7 ft above an open-grated floor. This positioning pro- vides the space that wiil be needed for the location of the fuel fill-and-drain tank, prepower and post- power sampling systems, fuel recovery tank, nuclear instrumentation, and enriching equipment. The off-center location alse serves te minimize the length of the NaK piping. The NaK aond off-gas piping connected to the reactor will pass through a thimbie-type passage or bulkheod with a bellows-type expansion joint in the double-walled cell. The opening wili be covered with a conical thermal sleeve which will be welded to the pressure cell wali. The piping 43 UNCI_ASSIFIED CRML - LR-- D5 447154 - L - REMOVABLE AT BROOF PLUGS ‘—-—7 S e - e 158 ft 9in T . - N ! LA - 2 P T o F’———;";;—;mfl \ll--;inng/ - :wL( Jm; = T e e ey . - GO i ] i stammaL o CODLOCKER INSTALLATION INSTALLATION OPERATING OPERATING - S : ENGINEERS ENGINEERS CREW CHIEF CREW | / HEALTH GRAFT FOREMAN | A - 5 o — I ' PHYS!ICS FiELD ENGINEERS ) ’ ' ’ T | e GORR:DOR powh: h . ~ - ) ' i i [ ——] - = T "~ ; ! = 1 %_j . 1 \\/, o2 1 | | = f H X INSTRUMENTATION S ‘ N ‘ T emop CONTROL ROOM ___J SERVICE | e i // i _ - £ !, 5 L 5 > o o ; i ! ! ‘ i o \ 6 i ‘ | b | I 1 [ Co : } > - i Ij HOUSE T 3 — 1 — — — A~ ROCF : T T R T T T T iy R h T e e , ; i T : | 5 — 1“ ‘ . REMOVABLE ROOF PLUGS Do PRESEN' i | ; MG HOUSE ,I j o~ ROOF ! i , / P NEW SWITGH | : NEW BLOWER HOUSE - : HOUSE ROGF : RO(I)F ‘ l i : f ] ; i . - i__, B ———— FIRST FLOOR PLAN Fig. 19. Plan of the ART Building. UNCLASSIFIED ORNL-LR-DWG 44724 . PENT OUSE SPECTRGMETER TUNNEL - NOTE: FOR PLAN OF ART BLILEDING SECTION A4 SEE Fig. 49. SECTION B-8 T P e - e {58 G 1 : : ;1 STEEL STACK-—-J B | g 4244 1tin. : " VENT ,\;/‘—k\ ! HOUSE— [/ L { | G ' - . . o Loy ] R METAL DUCT - _ ) Ei. BEBTY Sin, L J~RADIATOR o i T = PIT TS L ‘“m—_:fi }L = SECTION 0-0 e 10 20 30 SECTION ¢-C W Fig. 20, Section of the ART Building. Sy ORNL - LR-DWG 15056 . TOR ASSEMBLY CELL REAC @ Fig. 21. Reoctor Assembly Cell 46 will be anchored to the thermal sleeve., The annular space between the NaK and the off-gas piping end the thimble-type passage will be filled with insulation. Auxiliaery service pipes and tubes connected to the reactor will pass through three junction panels located on the west side of the cell. The openings will be covered with stiff plates which will be welded to the pressure-cell wall, The bulkhead between the double-walled cell is a bellows-type seal. Five doubly sealed junction panels for controls and instrumentation have been installed through the tank below the building floor grade on the east side of the tank as a part of another bulkhead system to pass the wires, pipes, tubes, etc,, required for the circuits and systems, The velume within the instrument and electrical wiring bulk- heads will then be maintained at a pressure of 200 psig to prevent outleakage from the inner tank, even in the event of a disaster. The various thermocouples, power wiring, etc., will be installed on the reactor assembly in the shop and fitted with disconnect plugs so that they can be plugged into the panel in a short period of time after the reactor assembly has been lowered into position in the test facility. This will minimize the amount of assembly work required in the field. Manholes 5 ft in diameter have been installed in the upper portion of the containers. The manhole in the water tank is located just above the flange on the inner tank to allow passage through both container walls and thus provide an entrance to the inner tank for use after placement of the top. Sufficient catwalks, ladders, and hoisting equip- ment will be installed within the inner tank to provide easy access for servicing all equipment. The control tunnel surrounds 180 deg of the north side of the cell, with all junction oanels exiting into the control tunnel which extends to the auxiliary equipment pit (formerly the ARE storage pit), The pit and the adjoining basement area will include such items as the lubricating oil pumps and coolers, hydraulic oil pumps, relays, switch gear, voltage regulators, and auxiliary equipment control panels. The Shielding Experiment Facility Tests made at the Tower Shielding Facility indicated that provision should be made for the measurement of the gamma-ray spectrum of the ART as a function of the angle of emission from the reactor shield, Therefore five collimator tubes have been provided to collimate four beams ra- diating from an equatorial point at the surface of the water shield at angles of from 0 to 70 deg from the radial direction and one beam from an equatorial point at the surface of the reactor pressure shell. The latter beam will be used only during low-power operation. The layout required for providing these beams is shown in Fig. 22. In addition to the facilities shown in Fig. 22, ¢ gamma-ray dosimeter will be located on the roof above the reactor. 47 8y “LABGRATORY UNCLASS FIED ORNL--LR - DWG 17125 -~ SPECTROMETER TUNNEL N :\\w SN HATCH B~ PUMP PIT~_ .‘ CONTROL TUNNEL 4 7 ................... Y . P PIT ... ADSORBER PIT-" CONTROL TUNNEL, RADIATOR AND PUMP PIT BLOWER HOUSE & B | 8 o 8 16 4 32 40 } e e § SCALE M FERY W _ —— - o PART BASEMENT PLAN Fig. 22. ART Shielding Experiment Facility. Part [ DESIGN AND DEVELOPMENT STUDIES DESIGN AND DEVELOPMENT STUDIES FUEL BDEVELOPMENT A thorough survey of materials that oppeared to be promising as heat transfer fluids was presented in ORNL-360.! The first requirement is that the fluid must be liquid and thermally stable over the temperature range from 1000 to 1800°F. A melfing point considerably below 1000°F would be prefer- able for ease in handling, but a substance that would be liquid at room temperature would be even better. Other desirable characteristics are low neutron absorption, low viscosity, high volumetric specific hect, and high thermal conductivity. Above all, it must be possible to contsin the liquid in a good structural material at high tempera- tures without serious corrosion or mass transfer of the structural materiol, If the heat transfer fluid is to serve also as a circulating fuel, it must be a suitable vehicle for uranium; that is, the solubility of uranium and its effect on the properties of the fluid must be considered. To be suitable for a high-temperature, liquid- fueled, epithermai reactor, a fuel sysiem must meet several stringent requirements. The liguid must contain a sufficient concentration of a uranium compound for a critical mass to be provided in the core volume and must melt at o temperature sub- stantially below the heat exchanger outlet tem- perature. [t must consist of elements of fow ab- sorption cross section for thermcl neutrons and must consist solely of compounds which are thermally stable at temperatures in excess of the core outlet temperature. A high thermal coefficient of expansion is desirable as an aid to selfregulation of the power level. In addition it must be stable to the intense radiation field ond must tolerate the fission process and the accumulation of fission products without serious adverse effect on its physical, heat fransfer, and chemical properties. Of the many fluid materials which have been con- sidered, molten fivorides are the only fluids which seem to be generally suitable as aircraft reactor fuels;2 g number of fluorides of low cross-section eiements appear to be particularly promising.? A, s, Kitzes, A Discussion of Liquid Metals as Pile Coclants, ORNL-3460 (Aug. 10, 1949). ZW, R, Grimes and D. G. Hili, High Temperature Fuel System, A Literature Survey, Y-657 (July 20, 1950) W. R. Grimes et al., p 915 in The Reactor Handbook, voi. 2, sec. 6 {1953). A mixture of NaF, ZrF , and UF, essentially a solution of Na,UF, in NaZrF,, containing 5.5 mole % of UF,, proved to be adequate as fuel for the ARE. Mixtures of this general composition are relatively noncorrosive to low-chromium nickel- base alloys and would appear to be adequate for use in the ART, but the melting point, vapor pressure, and heat transfer properties are irferior to those obtaincble from some other fluorides. An improvement in melting point and a slight improvement in vopor pressure and viscosity can be obtcined by the addition of REF to the NafF- ZrF ,-UF, system. However, the benefits obtained would seem to be marginal, considering the high cost of RbF and the added complexity of the system, unless very high uranium concentrations (above 6.5 mole %) are required for criticelity. Substantially lower melting points and vapor pressures con be obtained at some loss in heat transfer properties and at o slight increase in corrosion and moss fransfer rates with mixtures of NaF-BeF,-UF, as fuel. Some improvemen® over this mixture results if LiF-BeF,-UF, is used. in view of the diminished heat transfer performance, the toxicity of BeF,, and the cost of the required Li7, neither mixture appears to offer any real advantage over the NaF-ZrF -UF , system for air- craft reactors, as presently conceived. A slight improvement in melting poinf and a very real advantage in vapor pressure and heat transfer performance result from a solution of UF:4 in the ternary eutectic of NaF-KF-LiF. Although such mixtures are quite incompatible with Inconel and similar commercial chromium-bearing nickel alloys because of rapid mass transfer of chromium, the advantages will almost certainly justify the cost of the required Li7F if an adeguate structural and container material can be found. While information on the radiation behavior of several of these general classes of fiuoride fuels is meager, apparently there is no substantial dif- ference among them in their stability to the radia- tion fields and no one of them will show @ pro- nounced advantage in response to fission or fission-product buildup during reactor cperation. If an adequate container material becomes avail- able, it is very likely that the NaF-KF-LiF-UF fuef system will be chosen for use in sircroft reactors. As long, however, as [nconel or some similor chromium-bearing alioy of nickel has to be 51 used as the container, the NaF-ZrFA-UF4 mixture {or some slight variant of it) will continue to be preferred, The physical properties of the NaF-ZrF UF, (50-46-4 mele %) fuet are as follows: 525°C (977°F) 1230°C (2246°F) Melting temperature Boiling temperature Heat capqciiy3 {cal/g*°C) Liquid (550 < ¢ < 850°C) 8.26 1 0.03 Solid (350 < t < 500°C) 0.2% £0.03 Heat of fusion (cal/g) 57 * 10 Thermal conéucfivify4 1.3 10,2 [Bsu/hrft 4 OF /51)] Viscosity® {centipoises) At 600°C 8.5 At 700°C 5.4 At 800°C 3.7 Density® {g/em?) At 530 CrF, + Fe® 2FeF, + 3Cr° —— 3GrF, + 2Fe® 2CrF3 + Cr® —— 3CrF, Oxide films on the metal walls react with the fuel constituents (ZrF, or UF,)} fo yield structurdl metal fluorides: ZNi0 + ZrF4—=-——=-9- NiF, + Zréfi)2 2F6203 + 3ZeF , ——> 4FeF, + 3Zr0, 2Cr203 + 3ZrF ——> 4CrF 4 + 3Z¢0, These structural metal fluorides are then available for reaction with chromium, as shown above, It is, accordingly, necessary that the fluoride mixture and the Inconel be of especially high purity. If the purity specification is met and UF, is the uranium compound used, the reaction UF, + Cr° === CrF, + 2UF, becomes the rate-determining reaction in the cor- rosion process. in the thermal-convection loops, reduction of impurities in the fluoride mixture and equilibration of the metal surface with the UF, seem to require 200 to 250 hr and to produce void formation to o depth of 3 to 5 mils. In the forced-circulation loops, in which the flow rates of the fluorids mix- ture are much higher, these reactions proceed more rapidly but do net cause ony greater corrosion. For a given concentration of impurities of UIZ4 the depth of attack varies with the ratio of surface area to fuel volume; and, if equilibrium is estab- lished isothermally, the corrosion is reascnably uniform. However, in a system with a femperature differential the hotfest zone is preferentially attacked, When fetravalent uranium is present, the reaction UF, + Cr® == CrF, + 2UF, is responsible for some ‘‘mass transfer’’ in addition to the corrosion described cbove. In the mass transfer process, chromium removed from the metal walls in the hot zone is deposited on the metal walls in the lower temperature regions of the sys- tem. Since the equilibrium constant for the reaction is temperoture dependent, the reaction proceeds slightly further to the right in the high-tempercture zone {1500°F) then in the [ow-temperature zone. The UF; and the CrF, are soluble and therefore move with the circulating stream to the cooler zene, where a slight reversal of the reaction occurs and chromium metal is formed. This type of conversion is not readily apparent in thermal-convection loops operated for 500 hr, because in the first 500 hr of operation the mass fransfer effect is masked by the effects of impurities ond nonequilibrium con- ditions. [n loops operated for 1500 hr, however, the moss transfer effect is observable. It is esti- mated that, in Inconel systems circulating a zir- conium-base fluoride fuel mixture, corrosion by the mass transfer mechanism will increase by about 3 mils per 1000 hr. Data from a loop oper- ated for over 8000 hr substantiate this view; the total depth of attack was 25 mils, with a peak Inconel temperature of 1500°F, 33 The corrosion attack in the ART will be greatest in the hottest portion of the fuel circuit, and aofter 1000 hr of operation the attack is expected to be about 8 to 10 mils if all the uremium is present originally as UF .. If a mixture of UF, and UF, is used, the corrosion of Inconel by the fue! will be decreased., For instance, after 500 hr of circu- lation of @ UF ;-containing fused salt, attack of 1 to 2 mils is found, in contrast to the usual 3 to 5 mils. Experimenis are under way for defermining the proper UFB«*M«UF4 ratic for achieving minimum cotrosion along with adequate solubility in the zirconium-base fluoride mixture. The UF. is not sufficiently soluble in the NoF-ZrF, carrier to provide a critical mass, ond plating-out of the uronium weuld also be a problems it has been established that the zirconium-base flucride mixture fuels and Inconel will be com- patible under ART operating conditions for the 1000-hr expected life and that the ottack will not seriously weaken the reactor structure, but there is still some concern about the corrosion of the thin-walled (0.025-in.-thick} heat exchanger tubing. The cmount of mass transfer of chromium to the cold leg will be so small that there will not be an increase in pressure drop or ¢ decrease in heat transfer performance. The materials compatibility problem has also been investigoted for the materials of the reflector- moderator system in which sodium will flow in direct contact with both berytlium and Inconel. fn such @ system both temperature gradient mass transferand dissimilar metal mass transfer between the beryilium and Inconel can oceur. On the basis of numercus whirligig, thermal-convection loop, and forced-circulation loop tests conducted on beryllium-sedium-inconel systems, the former does not appear to be a serious problem if the tempera- ture is kept below 1300°F. The temperature gradient mass transfer detected on the cold-leg walls of any thermal-convection loop with a beryl- lium insert in the hot leg operated at 1300°F (cold leg, 1100°F) for periods of 1000 hr has been less than 200 ug per square centimeter of beryilium, The mass fransfer of Incenel by the sedium at temperctures below 1300°F is not considered to be a serious problem, A beryllium-sodium-Inconel forced-circulation loop in which egual creas of inconel and beryllium were exposed to the sodium showed no incresse in mass fronsfer over that found with ali-Inconel loops. This loop was oper- ated ot ¢ hot-leg temperature of 1300°F for 1000 hr, 54 and a 2-mi! deposit of crystals was found in the coid leg. The crystals were nickels=rich and con- tained little or no beryllium, Dissimilor metal mass transfer, in which the socdium ccts as a carrier between the beryllium and Inconel, hos been found to be o function of the separation distance between the two materials. When the beryllium and Inconel are separated by distances of greater than 5 mils, no continuous layers of nickel-bery{lium compound are formed on the surface of the Inconel in 1000 hr at 1200°F. Under these conditions a fine subsurface pre- cipitate, probably BeNi, forms to a depth of approx- imately 2 mils with ¢ 5-mil separation between the beryltium and the Inconel and to a depth of 1 mil with a 50-mil separation. The major compatibility problem in the reflector- moderator system will occur in those areas where inconel and beryllium are in direct contact. In 1000 hr ot 1200°F o 5-mil layer of brittle compound (4.5 mils of Be, Ni.; 0.5 mil of BeNi} formed on the Inconel when it was held in direct contact with beryllium with no applied pressure. Approxi- mately 1 mil of Inconel was consumed in the forma- tion of this layer. At 1300°F under a contact pressure of 500 psi a 25-mil layer of Be, Ni, + BeNi, which consumed 4 to 5 mils of Inconel, was formed on the Inconel in 1000 hr. One pessible sclution te this problem may be to chromium-plate the Inconel. Tests under the above conditions {1300°F, 500 psi, 1000 hr) indicate that a S-mil chromium plate can reduce the reaction to a cone- sumption of 2 mils of Inconel. Since the siructural metal of the fuel system will also be in contact with the sodium in the NaK in the heat exchange systems, intensive studies are under way on the competibility of inconel and sodium under dynamic conditions at ¢ peak sodium temperature of 1500°F. It is believed that mass transfer in these systems can be kept fo a tolerable minimum by the use of very pure, non-oxygen- containing sodium and very clean metal surfaces. In particular, it is expected that the beryllium surfaces will remove the oxygen impurity from the sodium and thus contribute to the reduction of mass transfer in the sedium system. Other structural materials have been studied or are being studied in an attempt to find o material superior to Inconel for subsequent aircraft reactors, The stainiess steels appear to be superior in sodium, but they are definitely inferior in the fluoride mixtures. Hastfelloy B is far superior in the fluoride mixtures, but it has poor fabricetional qualities. Various modifications of the basic nickel-molybdenum alloys are now being investi- goted. RADIATION EFFECTS ON STRUCTURAL MATERIALS The effects of irradiation on the corrosion of Inconel exposed to a fluoride fuel mixture and on the physical and chemical stability of the fuel mixture hove been investigated by irradiating Incenel capsules filled with static fuel in the MTR and by operating in-pile forced-circulation Inconel loops in the LITR and in the MTR. The relatively simple capsule tests have been used extensively for the evaluation of new matericls. The principal variables in these tests have been flux, fission power, time, and temperoture. In a fixed neutron flux the fission power was varied by adjusting the uranium content of the fuel mixture. Thermal- neutron fluxes ranging from 1017 to 104 neu- tronsscm=2esec—! and fission-power levels of 80 tc 8000 w/cm?® have been used in these tests. Almost ail the capsules have been irradiated for 300 hr, but in some of the recent tests the irradia- tion period was 600 to 800 hr. After irradiation the effects on the fuel mixture were studied by meas- uring the pressure of the evolved gas, by determin- ing the melting point of the fuel mixture, and by making petrographic and chemical analyses. The Inconel capsule was also examined for corrosion by standard metallographic techniques. In the many capsuie tests made to date no major changes that can be attributed to irradiatior, other than the normal burnup of the uranium, have occurred in the fuel mixtures. However, the anclytical method for the determinction of chromium in the irradiated fuel mixture is being rechecked for accuracy. The metaliographic examinations of Inconel capsules tested at 1500°F for 300 hr have shown the corrosion to be comparable to the corro- sion found under similar conditions in unirradiated capsules, that is, penetration to a depth of less than 4 mils. In capsules tested at a temperature of 2000°F and above, the penetration was tc a depth of more than 12 mils and there was grein growth. Three types of forced-circulation in-pile locps have been tested. A large loop was opercted in a horizontal beam hole of the LITR. The pump for circulating the fuel in this loop was placed outside the reactor shield. A smaller loop, including the pump, was operated in a verfical positien in the lattice of the LITR. A third loop was oserated completely within a beam hole of the MTK., The operating conditions for these loops are presented in Table 6, and results of chemical onalyses of the fuel mixtures circulated are given in Table 7. The LITR horizontal loop operated for 545 hr, cncluding 475 hr ot full reactor power, The loep generated 2.8 kw, with a maximum fission power of 400 w/em®, The Reynolds number of the circulated fuel was 5000, and there was o temperature differ» ential in the fuel system of 30°F, The voiume of TABLE 6. OPERATING CONDITIONS FOR {HCONEL FORCED-CIRCULATION IN-PILE LOOPS Operating Varicbles LITR Horizontel LITR Vertical MTR In-Pile Loop Loop l.osp No. 3 Na FoZrFA-U F, composition, mole % 62.512,5-25 63-25-12 53.50406.5 Maximum fission power, w/cm’ 400 500 730 Total power, kw 2.8 5.0 29 Dilution factor 180 10 3.5 Maximum fuel temperature, °F 1500 1500 1500 Temperature differenticl, °F 30 71 175-200 Reynclds number of fuel 5000 3000 5000 Operating time, hr 645 130 462 Time at full power, hr 475 30 271 Depth of corrosion attack, mil 1 i 1 55 TABLE 7. CHEMICAL ANALYSES OF FUEL MIXTURES CIRCULATED IN INCONEL FORCED-CIRCULATION IN-PILE LOGPS Loop Designation Sample Taken Minor Constituents (ppm) fron Chromium Nickel LITR horizontal loop Before filling 80 + 10 10t5 200 £ 100 After draining 180 T 40 150 + 10 0ts LITR vertical loop Before filling 20 10 80+ 10 145+ 20 After draining 37¢ 1 20 100 *+ 2¢ 50 £10 MTR inepile loop No. 3 Before filling 40 £ 10 60 £ 10 40 + 10 After draining 240 + 20 50 £ 10 100+ 20 the loop was large, and therefore there was a large dilution facter, that is, the ratio of the total system volume to the volume in the fissioning zone. Metaliographic analyses showed that there was less than 1 mil of corrosion of the Inconel walls of the loop. Chemical analyses showed that the irradiated fuel mixture contained 200 ppm or less Fe, Cr, and Ni. Therefore there was ne evi- dence of accelerated corrosion in this experiment, The LITR vertical loop cperated for 130 hr, with only 30 hr of the total operating period being at full reactor power. The loop generated 5 kw, with a maximum fission power of 500 w/em3. The Reynolds number of the fuel was 3000, and the temperature differential was 71°F., The surface- to-volume ratic was 20, and the dilution factor was 10. Chemical analyses showed that the irradiated fuel contained 370 ppm Fe, 100 ppm Cr, and 50 ppm Ni, Metallographic analysis of the loep showed that there was less than 1 mil of corrosion at the curved tip. The horizontal loop inserted in the MTR (MTR in-piie loop No. 3)7 operated for 462 hr, including 271 hr at power. The locp generated 29 kw, with a maximum power of 730 w/ecm3. The Reynoids number of the fuel was 5000, ond the temperature differential was 175 to 200°F for 103 hr and was T00°F for 168 hr, The dilution factor was about 3.5. Chemical analyses of fuel from MTR in-pile loop No. 3 showed that it contained 240 ppm Fe, 50 ppm Cr, and 100 ppm Ni. The high iron concentration wos prebably caused by o sampling difficulty. Examingtions of unetched metallographic sections °D. B. Trauger et al, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 27. 36 of the lnconel tubing showed that there was no corrosion penefration; etching revealed no attack tc a depth of more thon 1 mil. A slight amount of intergranular void formation was noted hut was neither dense nor deep. Measurements of wall thickness did not reveal any variations attributable to corrosion. The loop was examined carefully for effects of temperature variations between the inside ond outside walls of tubing at bends, but no effects of overheating were observed. Samples taken from the inlet, the center, and the exit side of the loop had less than 1 mil of corrosion. The low corresien is credited to the careful temperature contrel of the salt-metal interface and to the maxi~ mum wall temperature being below 1500°F at ail times, Future loops will be cperated at higher fission powers and therefore greater temperature differ- entials. The dilution factors will be kept low. New fuels and new alloys are being censidered for testing in future loops. STRUCTURAL DESIGN ANALYSES The Reactor The five principal probiems involved in the structural design analysis of the ART are the selection of the operating conditions for the reac- tor and the definition of the accidents or failures that might occur for which corrective action would be possible; the determination of the mechanical or pressure loads set up by these various operat- ing conditions; the calculation of the correspond- ing temperature distributions throughout the reac- tor; the selection of suitable design criteria; and the detailed stress analysis and structural design of the system, :{1:3;-;-:-:-:‘-,:"‘ The proposed operating program for the ART is based en the reactor being at temperature {(1200°F or greater) for 1500 hr. During the last 1100 hr of this time, the recctor will be critical and will be subjected to 25 full-power cycles to simulate flight requirements. In general, the most severe steady-state pressure loads to which the internal structure of the reactor will be exposed will occur during full-power operation, '% which is referred to ' and which serves as the "‘design-pcint condition as the basis for steady-state-load design analysis. Design for transient loads is bused in large part on the power-cycling conditions mentioned above and on various unscheduled changes in power level that will occur as corrective or safeguard action in the event of certain accidents. Not all the pos- sible operational situations which may arise during the test of the ART have yet been examined from a design viewpoint, but it is believed that those situations which will impose the most severe re- quirements on the structural design have already been considered. However, if further studies indicate the contrary, modifications will be made in the design or constraints will be imposed on the operational procedure. Information on tem- percture and pressure transients to be expected from off-design operation will be obtained frem tests performed on the ART simulator {(e.g., abrupt stoppage of several pumps). The pressure loads within the reactor have been determined for the design-point condition and for a condition involving the failure of ene fuel pump. It is believed that these two situations represent the most severe symmetric and unsymmetric in- ternal leads to which the reactor will be exposed. The calculation of the one-pump-out condition was performed in order to determine the stability of the reflector-moderator assembly. Under this circum- stance there would be no presswe drop through the fuel-to-NaK heat exchangers in the circuit supplied by the pump thet failed. This conditien would result, then, in a net side load on the re- flector of 12,400 ib. The reflector support ring has been designed to be stable against the re- sultant overturning moment, and the assembiy can be expected te remain seated against the north- head structure. 0The main heat exchanger headers and the core shells ore two important exceptions. The most severe loads on these members will occur in the event of the stoppage of a fuel pump. Feor design-point operation the pressures in the fuel and in the sodium circuits will be symmetric with respect to the vertical axis of the reactor; thus all resyltant forces on the principal struc- tural components (the reflector, north head, island, pressure shell, and pressure-shell liner) will be directed along this axis. In oddition to these vertical forces, horizontal reactions will ocecur between the reflector shell and the pressure-sheil liner and between the heot exchanger tubes and the reflector shells and the pressure-shell liner as a result of pressure differences and differential thermal expansion. In the heat exchanger tubes and the thin core and reflector shells these forces can cause large deflections and possibly buckling. The principal forces which must be accommo- dated at the design-point condition are the vertical loads thot will be imposed by the fuel. These loads consist of three components: forces result- ing from pressure drops in the fuel passages, weight forces, and buoyant forces. In the reflec- tor-moderator assembly the pressure-drop force will be 58,000 ib, the weight 3,000 Ib, und the buoyant force 4,700 [b. The net force wili be up- ward and will press the reflector against the north head, which, in turn, will transmit the force to the pressure shell, There will also be an upward thrust of 20,900 Ib on the north-head region of the pressure shell from the combined action of pressure, weight, and buoyant forces on the heat exchanger. The heat exchanger and reflector forces will be brought into balance through the action of corresponding pres- sure forces on the inside of the pressure shell (transmitted by the liner), which will force the pressure shell downward (Fig. 23). Similar buoy- ant, weight, ond pressure-drop forces will alseo act on the island assembly, but these forces will be relatively small and can be carried by the shell structure without the aid of special structural members. Caleulations are being made for determining the three-dimensional temperature distributions through- out the sodium circuit, the fuel, and the principal structural members for the various operating condi- tions of the reactor. In the initial design studies, thermal loads and stress values were used that were based on very preliminary estimates of the temperature structure, and therefore the very de- tailed analysis being made will serve as ¢ check on the ceolant-flow provisions: |t will also pro- vide more accurate estimates of the thermal-stress distribution in critical members. 57 85 NORTH HEAD 7/ [y \ o7 TN i {"'fz J/ [ \'L VoA “\;"\ ““\ ! : 3 | b “.\ P ! ;" ! JH‘ f, By \\; vy / r/ { i}.- N N . UPTHRUST FROM REFLECTOR (59,656 Ib) . FUFL PUMP DISCHARGE PRESSURE FORCE {340¢€ Ib} . FUEL PUMP SUCTION PRESSURE FORCE (7747 Ib) . SODIUM PRESSURE FORCE ON NORTH HEAD {psi) . NORTH-HEAD WET WEIGHT {1000 Ib) . PRESSURE SHELL REACTION LLINER G =T X & H . HEAT EXCHANGER PRESSURE DROP FORCE (24,216 ib) . FUEL PRESSURE FORCE FROM CORE EXIT (65,264 b} . STATIC HEAD (B343 \b} . SODIUM PRESSURE FORCE . LINER WEIGHT (1000 b) . ISLAND (200 ib) . PRESSURE SHELL SUPPORTING REACTOR {J+98,993 ib) Fig. 23. Pressure L.oad Distributions ot Full Power. »w o O T o GSESRET ORN|-LR—DWG 16018 PRESSURE SHELL . NORTH-HEAD RZACTION {68,809 ib) . LINER LOADS (M)} . MEAT EXCHANGER LOAD (20,912 Ib} . PRESSURE SHELL WEIGHT (3500 ib) . PUMPS (4200 Ib) . SUPPORT LOAD ON REACTOR {(REACTOR WEIGHT = 44,000 ib) Precise calculations of the temperature profiles throughout the reactor require detailed information on the energy deposition from radiation and neu- tron reactions and on the temperature structure along the boundaries of the fiuid passages. In- formation on the energy depesition is to be ob- tained by two independent methods: a buildup- factor technique in preparation at Pratt & Whitney and a Monte Carle method developed at ORNL. The nuclear reactien dats required for these pro- grams are based on the two-dimensional multigroup flux calculations suvpplied by the Curtiss-Wright Corporation. The dual program has been initiated because the Monte Carlo method, althcugh more accurate, will be several months in preparation, and it is believed that the relatively quick Prott & Whitney method will suffice for preliminary three- dimensional data. Finally, @ more accurate picture of the temperature profiles through the fuel will be obtained from analyses of the results of the high-temperature critical experiment and of the tests of the half-scale model of the ART core, which wutilizes a velume heat source to simulote the fission heating. The utilization of the tempercture and load dis- tribution information in the design cnalysis of the reactor system poses some difficulties in a broad technological sense because of the lack of an established design philosophy for high-temperature operation. In general, relatively little design ex- perience has been acquired in providing for the effects of thermal cycling, strain cycling, creep buckling, and thermal relaxation. Suitable design criteria are being formulated for the ART by combining informotion obtained from mafterials testing programs, component testing programs, and the Engineering Test Unit (see Part 1V}). In the materials testing programs, data are being obtained on the properties of Inconel and beryllium at the temperatures of interest in order to acqguire insight intc the nature of the basic thermal phenomenon involved. Data on the creep and tensile properties of the two materials, their behavior under strain cycling, and their relexation properties are of particular interest. The creep and strength data!! are re- quired for the design of members that will be sub- jected to continucus loads over prolonged periods _of time {e.g., loads resulting from operating pres- ViThis work is under way at ORML and The Brush Beryllium Co. The strain-cycling data'? are being cor- 13 and sures). related according to the Coffin formulation will serve as the basis for the design analysis of structures which will be repeatedly subjected to mechanical or thermal loads that will produce plastic deformation in the material. The reloxation data'4 are to be used mainly to determine the amourt of plastic strain developed under cyclic loads. The creep-buckling information!® is re- quired for the design analysis of the various core and reflector shells. These shells will be sub- jected to pressure differentials and to temperature gradienits both through their thickness cnd along their surfaces. In some instances the matericals test information obtained from the programs mentioned above may be used directly to estimote the thermal deforma- tion and buckling characteristics of components and to determine the life of parts subjected to cyclic loads. However, this direct application wili be effective only for relatively simple struciural configurations and cannot be used for many of the important structural members of the reactor. In the latter cases it will be necessary to resert to component or scale-mode! tests under operating conditions. For example, a test is under way on ¢ one-fourth-scale model of the cuter core shell to determine whether the core shells will survive the thermal cycling to which they will be exposed during the operational life of the reactor. The eperational conditiens for this test were based on the anticipated program of the ART, and the tem- peratures and hold-times involved were defermined from relaxation and strain-cycling data on Inconel (Fig. 24). This test should indicate, also, the extent to which the strain-cycling and relaxation data based on simple uniaxicl stress conditions can be extropolated to the more complex patierns encountered in the actual design. The detailed stress analysis of the ART, now under way, consists in examining the system op- erating pressures and thermal and cyclic loads. The operating pressure loads were used to size the principal structural members. The criterion 1297his work is under way at ORNL and at the Uni- versity of Alobama, 13, F. Cofttin, Trans, Am. Soc. Mech. Engrs. 76, 931950 (1954). "“To be supplied by ORNL, WADC, and the Uni- versity of Michigan. 15Bata now being collected at Pratt & Whitney and the University of Syracuse. 59 SECRET ORNL-LR—DWG 16049 25,000 e ——— — it — - TOTAL STRAN 0.3 % (og = 54,000 ps) : ; 20,000 - - R e L e F—TOTAL STRAIN 0.3 % (o, = 54,000 psi} | | | : . ‘ i 15,000 i _ e el e e h_.+___. N ; B FULL POWER i\ e HOLD | I« L : ! 10,000 M R . o e e | _~TEST CYCLE -~ ART CORE SHELL [~ z L - 5,000 \\ i,‘:I"ffgo TQ 60 Mw N - i’M e o sy ' a he- HEAT-UP w O bF——""0- 0 TO 10 Mw —— - —-- - —_—b - ; ! ‘ | - COOL DOWN 5,000 |—— #‘"",Q;QO&,QQWN o _ o ! -—‘-_—-_4 % . // - | 40,000 |-~ -+ i Lo T : ——— - A~ 1T ISOTHERMAL _ | [1SOTHERMAL o o -15,000 | _A+JIF S : . R L e b 1 --H. -20,000 b—— ————— e e | e e e | -25000 — - --- e — gl Lo 0 1 2 3 4 5 6 7 8 9 10 1 16 17 18 19 TIME {hr) Fig. 24. Conditions for Thermal Cycling Tests of One-Fourth-Scale Model of ART Core Shell in Comparison with ART Thermal-Cyecling Conditions. presently in use requires that the stress level created by these loads be no more than one-fourth the value known to cause rupture from creep if applied for 1500 hr. The stress analysis is based on elastic theory, but for very complicated con- figurations the analysis is supported by experi- mental studies on modeis of the actual member. This scheme of analysis has been applied in a preliminary manner to all structural components of the reactor, its support structure, the NaK piping inside and outside the pressure cell, and the heat dump systems {NoK pumps and radiators). These preliminary studies have been completed, and attention is now being focused con the detailed stress analysis of the major structural members which are a part of the reactor proper; these mem- bers include the north-head double-deck composite structure, the reflector suppert ring, the core and reflector shells, ond the fuel dump tank and supports. A tentative set of design criteria has been se- lected for the detailed examination of the system from the viewpoint of thermal and cyelic loads. For members that will be subjected to thermal foads which may occur only a few times during their life, the design criterion is that the strains produced by these loads not exceed 0.2%. Based on the concept of an elastic stress-strain relation, this corresponds to a stress of around 30,000 psi at 1300°F. Such a criterion, although in agreement with the design philosophy of the ARE, is believed to be extremely conservative, and o more realistic one is being considered and will be evaluated by the results obtained from the strain-cycling fests for the lower number of cycles. The criterion for the design of members subjected to cyclic plastic strain is also based on the requirement that the maximum deformation during a cycle not exceed 0.2%. At temperctures of the order of 1300°F « specimen thus loaded should survive more than 200 cycles. This figure is based on the best strain-cycling data now available. ' This crite- rion will be used until more reliable information is obtained from the strain-cycling test program under way at ORNL ond ot the University of Alabama. The thermal-stress criteria outlined here are in use in the analysis of the cyelic heating of the core and reflector shells, the pressure-shell liner, and the thermal sleeve attachments for all pipe cutlets from the reactor, as well as in the thermal- stress analysis of the fuel pumps and of the NaK manifolding ot the reactor. An anclysis is alse in progress for determining the thermal expansions and distortions of the various shells and the beryllium within the reactor. These calculations require the temperature disiributions previoustly menticned, and the results obfained from the work will determine the final tolerances and clearances between these members. It should be recognized that the proper fit of the parts under the operating conditions will define the cold dimensions to which the system must be assembled. The cnalytical studies described above are sup- ported in many creas by paraliel programs of ex- perimental stress analysis. The bulk of the work is being carried out at the University of Tennessee and inciudes the octual model tests of the north- head composite-deck sfructure, the pump barrel and NcK pipe aftachments to the pressure shell, the main and auxiliery heat exchanger headers, the reflecior support ring, the blowout patch for the pressure shell, and the NaK piping systems out- side the reactor cell. The emphasis in the design analysis under way and that programed for the near future is placed on the re-evaluation of the design criteria, the de- tailed stress analysis of the primary structure, the completion of strain-cycling tests of component models, and the study of off-design and transient operating conditions. Auxiliary Components Several major structural compenents external fo the reactor, including the reactor support system, the fuel fill-and-drain tank and support, and the main NaK piping which carries the heat from the reactor to the radiators, have also been analyzed. Y nuclear Propulsion Program Engineering Progress Report No. 18, October 1, 1955—-December 31, 1955, PWAC-554. The reactor is to be suspended from an overhead bridgelike structure by means of the four pump barrels {(Fig. 25). The attachment of the individual barrels to the bridge aliows herizontal motion of the barrels so as to ccecommodate the relative thermal growth between the reactor {at operating temperature) and the bridge (at room temperature). Vertical motion of the barrels is completely re- strained. The bridge is fixed at each end fo © flexible column consisting of Il-in.-thick steel plates, 28 in. wide and 123 in. long. The [ocd carried by each column is approximately 45,000 1b. This value is somewhere between one-fourth and one-half the velue required to cripple the column, The principol function of the flexible celumns is to aliow complete freedom for the NalC lines to expand in going from room temperature to the operating temperciure of the reactor. During full- power operation the upper row of NaK lines will be at 1070°F, the lower row at 1500°F. This will result in o horizontal growth of 3{4 in. in the upper fines and '55’8 in. in the lewer lines. [f the reactor is mounted in the cold condition precisely cver the center line of the column bases, these expansions will translate and rotate the reactor out of the neutral position and thereby introduce bending stresses into the columns (Fig. 26). In crder to eliminate the bending stresses, it is planned te precut the NaK lines so that at room temperature the reactor wiil be lecated 1 in. off the neutral position toward the cell wall through which the NaK lines enter. As the reactor heats up, these fines will expand end move the reactor into the neutral position, thus removing the bending loads on the columns. Although the flexible columns can allow for the gress expansions of the NaK lines, they cannot provide for the differential expansion between the fines of any one row. In order te provide some margin for operational accidents and freedom in controlling NaK temperatures, it is plenned to add several bends into each line to accommadate 300 to 400°F temperature differences between adjacent lines. The piping layout proposed for the reactor celt for this purpose is shown in Fig. 27. The basic design features of the NaK piping out- side the reactor cell are similar to those inside the cell. The principal design reguirement is that the pipe supports and end aftachments have suf- ficient flexibility for the lines to expand in going from room temperature to the operating temperature without being subjected to excessive thermal 61 o . | -—m}, Lrw v | - - | S *Wil ER SHIELD CONTAINER \\ " FLEXIBLE COLUMN " | REACTOR \ ‘ ,r,u.__r__._ ][ _I;, - 5 ft 8% in—— - [ i 1 | ! / | I ! ! .f =t FONELCENTA ORKL~LR—~DWG 16123 i //// AT TEMPERATURE, WITHOUT PRECUTTING OF NoK LINES (o) 62 6|n ‘ —.18in. S S 6|n E —9ft Oin- - e 101 ;3 in. REACTOR CELL] -‘:_-—_—‘"‘“_:1:%— _____ ¢ S— NoK LINES 1 121t 8'9 in. 5” olyin Lt o _ Fig. 25. ART Support Structure. SONPITENTIAY. ORNL-LR-DWG 16124 _~BRIDGE ~~PUMP BARRELS - REACTOR "~ TN COLUMNS D AT TEMPERATURE,WITH WITH NaK LINES PRECUT PRECUTTING OF N¢K LINES (&) (c} Fig. 26. Movement of Reactor Due 1o Expansion of NaK Lines. POSITION OF REACTQR AT ROOM TEMPERATURE, SECEELL’ ORNL-LR-DWG #:022 CELL WALL ~__ | | o = o Fig. 27. Layout of NaK Piping in Reactor Cell. stresses and without producing excessive loads on the radiators and pump casings. One of the NaK radiator-pump assemblies in the ART system is shown in Fig. 28. The individual lines are welded at the reactor cell wall, and the joints may be considered as fixed points in the piping circuit. The radiator ends of the lines are welded to their respective radiators, and the entire pump-radiator assembly is suspended by a system of spring hangers from overhead sc as to allow freedom of motion in the principal direction of the NaK lines. This freedom allows for the over-all growth of the lines, and the various bends in the individual lines provide increased flexibility to accommodate dif- ferences in line lengths. The operation and flew characteristics of the fuel fill-ond-drain tank system are discussed in Part II. Since this tank is to serve as an eversafe depository for the fuel, it must survive some 2000 hr of operation at temperature and possibly 63 Eanrronmrat ORML- _R-D¥G 13202 gt sk in. e PR | [ i TN e e e = e f Y Fig. 28. NoK-to-Air Radiator, NaK Pump, and NaK Piping External to Cell. several fast dumps. The principal structural re- quirements are determined by the 40-psi NaK pump- ing pressure, which will produce creep in the tank structure. The design criteria, then, are based on creep-rupture and thermal-shock censiderations. The most sensitive areas of the design are the joints between the tube header sheets (Fig. 16) and the inner cylinder. The discontinuity stresses in these regions are of the order of 6000 psi, but the stresses are expected to decay rapidly, once the system comes to temperature, as a result of 64 - shell. the stress-relaxation phenomenon. Since this com- ponient is of vital importance to the over-all safety of the experiment, it is planned to test the entire tank system by using one of the NaK circuits, This test will determine the adequacy of the de- sign in withstanding the creep and thermal shock effects mentioned cbove. The support of the dump tank assembly is ac- complished with the «id of a nitrogen cylinder located beneath the tank (Fig. 29). Although the tank is oftached rigidly to the reactor pressure shell through the fuel drain line, the major portion of the tank weight is not ailowed to bear on the shell. The total weight of the tank, including the fuel, is approximately 6000 Ib, of which 5000 Ib This arrangement introduces some complication in re- gard to the stability of the support system, but an analysis has shown that the proposed arrangement will be carried by the nitrogen cylinder. of lever arms, the structural stiffness of the drain fines, and the weights are well within the stability limits of the system. RADIATICON HEATING ON THE ART EQUATORIAL PLANE IN THE VICINITY OF THE FUEL-TO-NaK HEAT EXCHANGER The radiation heating to be expected in the ART was calculated so as to provide a basis for the design of cooling systems. The results of the calculations of the radiation heating on the ART equatorial plane in the outer 3 cm of the beryllium reflector and in the Inconel and the boron-contain- ing shells on both sides of the fuel-to-NaK heat exchanger are presented in Figs. 30 and 31. The total gamma-ray heating in each region is given in Fig. 30, as well as the heating from the sources which are the main contributors to the total in each The encircled numbers on Fig. 30 refer to the sources described in Table 8. The data on heating in the copper-boron layer by alpha particles from the B1%(n,a)Li? reaction are plotted in Fig. 31. The heating goes to infinity at the face of the layer closest to the core because the heating at various points is governed by an E | function, ® dA o | e 2 1 A where A is the mean free path. The integral under the curve will be finite. SOMNFHDENTTAL™ ORNL—-LR-DWG 16127 SUPPORT BRiIDGE P \\\\y —— ] - ’/ % |: b, il d \\ . - DUMP VALVE VALVE ACTUATOR ———— 5., (NGK SUPPLY ? W - (/R ATTACHMENT - mome——=-= TO FLOOR Fig. 29. Fuel Fill-and-Drein Tank Support. 65 99 SRERE L . OR/NL-LR- DG 14916 10 ———y i - STAINLESS STEFL g 9 TOTAL FEATING. ——{0—— HEATING FROM SCURCE . {SEE FiUFL GARP - 1 eal, TAELE 1.2 1. @ HEATING FROM SOURCES 1,2, AND | " A ;i . . i & 0 - 3 (SEE TABLE 4.2.4). | ; 2 = _ ! & < % = ALl DIMENSIONS IN cm., . : L w L2 o - ; ] o = T —l @ i L b - o L o ; Lr;;‘ = d = E - = Pt i [¥a = 5 S | 7o = o 8. & 1. \ W 3 2 . ! % = ol O L = = = : =’ o g , - < = O : _ o 2 m & o x i ; 1 N = O i HETL A : ; o it = A HEL UM : ‘ > « x m ” == 30DIUY i I o GAP P i il 0 o o - m i ~ - - O ] P oA I T8 < -« o 457 T i 8 - 057 = 34 ; s o e o - 0457 = 0.318~ \ 5 - = - z ] e ol baCmd G ™ bl O P A e D = 2 S % Q S ] o= D] e ] 2.4 — T‘Z < i ,,; o L 3 & . = = — i T 5 1 i i O @ L o = = S E 3 1 = o S 4 — - T R F; o3 o o i T z w oy Ll i 7 o o l @ 3 L _ - 4 N b , .8 | ¢ ) : 1 | 1) L : STAINLESS STEEL 4~ ™1 \l’F‘. g i \‘:'A\ = = o I . ——- 41 . o 4\’—5‘ o g/ . e —— e - - _t:& . . o o~ =~ ' N AV G~ . — < \ o Gd - ® 2 (G 2 @\ N A8 //@ ®\ 55 v \ : e e~ - (G; Fig. 30. Gamma-Ray Heating in the Vicinity of the Fuel-to-NaK Heat Exchanger on the E quatorial Plaone of the ART. TABLE 8. SOURCES OF RADIATION HEATING CONSIDERED IN CALCULATING THE RESUL TS PRESENTED IN FIG. 30 Source Source No Source Strengih 1 Prompt gamma rays in the fuel region of the core of the reactor 28.3 w/cm3 2 Decay gamma rays in the fuel region of the core of the reactor 6.84 w/cm3 3 Gamma rays from inelastic scattering of neutrons in the fuel region of the core 10.1 w/cm3 4 Capture gammc rays in the cuter core shell 41.4 w/t:m2 5 Capture gamma rays in the reflector (average) 0.5 w/em® 6 Capture gamma rays in the first Incone! shell cutside the beryllium reflector 22.5 W/cm3 7 Boron capture gamma rays in copper-boron layer 1.8 W/t:m2 8 Alpho particles from the Bm(n,a)Lij reaction in the copper-boron layer (average) 42 w/cm:?' g Decay gamma radiation from the fuel in the heat exchanger 2.3 w/’c:m3 10 Gamma rays from inelastic scattering of neutrons in first 9 em of reflector 0.7 w/«:.m3 {avercge) 11 Capture gamma rays from delayed neutrons in the heat exchanger and Inconel C.1 w/t:'rn3 shells {including the pressure sheil) 12 Capture gamma rays it the copper of the copper-boron layer 0.5 W/cm2 13 Gamma rays from inelastic scattering in both core shelis 4 w/c:m‘2 14 Capture gamme rays in the island core shell 41.4 w/c:m2 *In Fig. 30 the data for hecting from scurces 1, 2, 3 are combined and labeled 4, The heating from sources 10 to 14 was neg- lected. Their combined contributions to the heat- ing in the region being considered was estimated to be about 5% of the total heating. 7 RADIATION HEATING IN YARIOUS REGIONS OF THE NORTH HEAD The radiation heating to be expected in various regions in the north head of the ART was calcu- lated in order to supply numbers from which ther- mal-stress calculations could be made. Because of the complexity and the time that would be in- volved in calculating accurately the heating in all the regions of the north head, it was decided to make preliminary estimates of the deposition rates. More accurate values calculated for other regions of the reactor were used as guides. In all cases the tendency was to overestimate the heating. Caiculations were made of the heat-deposition rate in a slab of Incone!l bounded on one side by 7 Eor details of these caleulations see Chap, 1.2 of ANP Quar. Prog. Rep. June 10, 1956, ORNL-2106, p 28. an infinite fuel region containing the scurces of radiation. This heat-generation rate was used in all regions in the north head which are bounded by finite fuel volumes. The heat-deposition rates in a slab of Inconel bounded on one side by slabs of sodium of various thicknesses were calculated, and the results were extrapolated and interpolated to obtain the heat- generation rates in the Inconel regions of the north head which are bounded by various thicknesses of sodium. Fairly accurate calculations were made of the heat-deposition rates in the Inconel filler plates below the island ond in the vicinity of the fuei-to- NaK heat exchanger on the equatorial plane of the reactor. These results were used as ¢ guide in estimating the heating in some north-head regions, and new values were obtained by compensating (by simple exponential attenuation) for decreased beryllium thicknesses, penetrations through addi- tional fuel layers, increased thermal-neutfron leak- age currents into the north head, etc. 67 SEGRET ORNL—LR—DWG 14317 CALCULATED HEATING {w/cm®) . _co020 ,-n_,|L,,, - al _L__ I L o o 0ca 0.08 Q42 Q4G Q.20 Q24 THICKNESS OF LAYER {(cm] Fig. 31. Heating in Copper-Beoron Layer by Alpha Particles from the B'%»,a)Li7 Reacticn. [t was assumed that a neutron current of 7 x 1073 =1 was escaping uniformly from the upper portion of the core and that 1 Mw of fis- sion pcwer was being generated in the fuel regions neutrons-cm-_ 2-s.ec of the north head by neutrons escaping into this region. The latter increased the gamma rays in the fuel by about 30%. The sources of gamma rodiation considered were those from the heat exchangers, the boron, the core shells, the beryllium, the Inconel sheli capture gamma rays, the sodium and fuel in the north head, and the fuel in the core. The sources of heta particles considered were those from the gases in the fuel-expansion tank, and the sources of alpha particles were tcken to be those from boron cap- The average values of heat generation ob- in these calculations are presented in tures. tained 68 Table 9. The configuration of the north head is shown in Fig. 32. BETA- AND GAMMA-RAY ACTIVITY IN THE FUEL-EXPANSION CHAMBER AND THE OFF-GAS SYSTEM The power-source distribution of the activity of the gases in the space above the fuel in the fuel- expansion chamber and in the off-gas line has been determined. The resuits obtained are to be used in the calculotion of the radiation heating and the thermal stresses in this region of the reactor, The radicactive constituents of the gas in this space will be the gaseous fission products, xenon and krypton, and their deughter products. There is also a possibility that some volatile fission- product fluorides will be formed in the fuel and will escape into this area. However, it has been shown '® that if all the fission-product fluorides entered this space they would add very little ac- tivity to that already caused by the gaseous fis- sion products and their daughters. Thus their effect has been neglected. Also, there is some question as to whether the daughter products of the fission gases will actually be carried down- stream by the off-gas system or whether they will be deposited on the enclosing walls as they are formed. In order te get a cecnservative estimate of the power-source distribution, it was decided to freat the daughter products of xenon and krypton as gases (except insofur as their purging from the fuel into the fuel-expansion chamber is concerned). The total power and the power density in the gas space of the fuel-expansion tank as o function of the volume of the gas and the helium tlow rate are given in Fig. 33. the very short- and very long-lived nuclides of xenon and krypton {along with their decay products) were neglected, Since the fuel circulation time in the ART will be less than 3 sec, nuclides with half lives less than this valve will decay mostly fn the calculation of the curves in the fuel before it reaches the purging pumps. Thus very few atoms with half lives of less than about 3 sec would get into the gas space. Also, for nuclides with long half lives {(greater than chout 100 hr), the number of disintegrations taking 18, 1. Newgard, Fission Product Activity and Decay Heat Distribution in the Circulating Fuel Reactor with Fission Gas Stripping, TIM-205 (Sept. 28, 1955). EECRET CGRNL-LR—-DWG §4947 ¢ EXPANSION TANK —EEN PRESSURE SHELL>/ , i P LINER— ey FUEL 0\ T EXPANSION | TANK ~® /@ i i - ¢ ISLAND Fig. 32. Configuration of ART North Head Showing Members Referred to in Table 9. 69 TABLE 9. AVERAGE HEAT GENERATION RATES IN MEMBERS OF ART NORTH HEAD Member Heat Generation No. * Description (w/cms) 1 Pressure shell (below sodium expansion tank} 4 2 Liner 6 w/cm® + 16 w/cm? on expansion- tank surface due to beta roys 3 Fuel expansion-tank baffle 3 4 Fuel expansion-tank wall 6 5 Upper deck {regions with sodium on both sides) 2 6 Upper deck {regions with fuel on both sides) 15 7 Swirl chamber baffle 3 8 Swirl chomber wall 3 g Lower deck (regions with fuel below and sodium ebove) 8 10 Lower deck (regions with fuel on both sides) 12 1 Copper-boron tiles 25 w/em2t + 6 w/emS, where t = thickness of tiles (cm)} 12 Fitler block 3 13 Beryilivm support struts 10 14 Filler block 1 15 Copper-baron tile 30 16 Flat section of lower support ring 15 17 Strut part of lower support ring 3 18 Lower support ring 1.5 *See Fig. 32 for location of member. S GRNL—-_R—DWS 14918 3N — e —————— - e e ; POWER DENSITY | ——— TOTAL POWCR } 20— e . © om0 RELIUM FLOW RATF, 1000 liters /oy (TP} [ 140 — - F;. D8 e N I Ce—— = - . b T e o 00 - i - & % - y - 2 n E Z 1 o yl = 2 & O = = o 100 200 302 400 800 600 7OC 820 YOLUME OF CAS 5PACE flf‘..’%) Fig. 33. Total Power and Power Density in the Gas Space of the ART as o Function of the Gas Volume and the Helium Flow Rate for a Fuel Flow Rate of 22 gpm. 70 place in the fuel-expansion chamber and off-gas line would be small, since the dwell time at the assumed helium flow rates is very short. There- fore these nuciides may be neglected. In this study 32 nuclides were considered, 16 being isotcpes of xenon and krypton and 16 being their daughter products. The main contributors to the power distribution are the daughter products and not the nuclides of xenon and krypton them- selves. In all cases the daughter products con- tribute about 50 to 0% of the total power distribu- tion. Of the total power, about 90% is due to the beta-ray decays, with only 10% being due fo gamma-ray decays. Thus in determining the heat- ing caused by these gases, it is seen that the heat deposition will occur mainly in a small surface layer of the materials surrounding the gases in the fuel-expansion chamber and the off-gas line. The power density in the off-gas line as a func- tion of time ond gas volume for helium flow rates of 1000 and 3000 liters/day (STP) is given in Fig. 34. The time axis can be converted into lengths along the off-gas line by dividing the volume flow rate of the helium gas by the cross- sectional crea of the off-gas pipe. gives the power-source density of 1 ecm3 of the Thus Fig. 34 These plots were made by using the well-known equations of the decay of parent products and the buildup of their daughters as a function of time. The initial conditions at the beginning of the off- gas line were taken as the equilibrium conditions gas ot any position in the off-gas line. that weuld prevail in the fuel-expansion tank. SECRET ORNL—~LR—DWG t43519 ~ T T ‘ I l N T L[] T T[T T S~ e e HELIUM FLOW = 3000 liters/day (STP) 5 ~| | ————— HELIUM FLOW = 10CO0 liters/day {STP) N ' L R 30 g -] S . bl ‘ i : | GAS VOLUME = 20in.3 | ! i f | | | . } ! l i | | aEil AN L | L —_— | | N , | Pl \ ‘anin3 | ! ' —_ ! e 200 5 "E A \P“T“ 3 i 3 | ‘ 3 T\\ \ | =20 e e j ] \ ot & = i l : \ \ _ | @ | | = i \ || E & | N = | | 5 N AU ; g | T e 3001R.% ! : e — ] & : L : ——l ‘ ; \\ P : : - . ! . C ooin3 | | | T14=.20in ™~ : | 10 t—_-_-_"'i’—""-i' __-__ - fi\_""_\‘ \x i ‘L L i - \ b.‘-hfi\ \ 1 300in | LT T - _—"'_"""‘""-!-u—‘»_._w T~ ~~ 1 i .--"'-..-.-\ I\E\\ : i _, 1‘ ™~ ™ _ ‘ . i i | - | =3 B ; ST T } M, ’ :\*\ [ ; \\ ! \*._:: | *"l—.____ " Q i L t e ! '_""‘—v—l 1 2 5 10 2 5 10 2 5 103 2 5 108 TIME {sec) Fig. 34. Power Density in the Off.Gas Line as a Function of Time and Gas Yolume in the Expansion Tank for o Fuel Flow Rate of 22 gpm. 71 Part IV ENGINEERING TEST UNIT ............. ENGINEERING TEST UNIT A full-scale zero-power engineering test unit (ETU) is being fobricated that is a prototype of the ART. Assembly of the ETU will provide a test of the feasibility and efficiency of the pro- cedures prior to assembly of the ART. Thus time-consuming and expensive rework operations such as those required for the ARE will be avoided. The first reactor assembly fabricated may have, for example, a number of doubtfu! welds that would not prevent the initiation of nonnuclear shakedown tests but would have to be reworked if the assembly were to be used as the high-power ART. The fabrication experience gained will be used in assembling the second unit. A second and even more vital reason for fabri- cation and cperation of the ETU arises from the complexity of the stresses on the assembly. The complex geometry, the wide variety of combi- nations of thermal and pressure stresses, and the difficulty of predicting the magnitude and direction of thermal warping and distortion make it essential to run a comprehensive test on a nonnuclear assembly, even though not all the conditions can be simulated. Experience at the Knolls Atomic Power Laboratory! has forcefully demonsirated the importance of such tests when dealing with a high-performance complex that is to operate under conditions for which there are little data or ex- perience. Upcn completion of the tests, the ETU will be completely disassembled and thoroughly inspected. Invaluable shakedown and endurance test ex- perience will be obtained, and the fraining of the setup and operating crews during such a test will expedite assembly and operation of the ART. [t will be possible, also, to obtain heat transfer information on the radiators and on the NaK-to-fuel and sodium-to-NaK circuits and to test some of the instruments to be used on the ART, The heat for the nonnuclear ETU will be sup- plied by two gas furnaces, which will replace the radiators in two of the four main NaK circuits. It was originally intended that the capacity of each furnace would be 5 Mw, but each furnace has been reduced to a capacity of 1| Mw because of procurement and instatlation difficulties. Along with the reduction in the heat input, the radiators TR. W. Lockhart et al., Review of SIR Project Model Steam Generator Integrity, KAPL-1450 (Nov. 1, 1955). have been eliminated from the other two man NaK circuits. Radiators will be included only in the reflector-moderator {sodium-to-NaK) cooling circuit. These radiators will permit a determinafion of performance characteristics of not only the radi- ators but alsc the louvers and the heat-barrier doors to be used with them. It is particularly important that the sensitivity of the controls for these units be determined at low loads. Since the units will be essentially the same as those to be used in the main circuits of the ART, the test of the reflector-moderator cooling circuit will serve tc answer meny basic questions regarding the performance characteristics of the main circuits on the ART. Since there will be no after-heat and ro radie- active off-gases, the fuel dump tank and the off-gas system, rather than being prototypes of those to be used with the ART, have been simplified to expedite fabrication, construction, installation, and testing of the ETU. The fuel enrichment, recovery, and sampling systems are not to be included in this nonnuclear test assembly, The ART design will be followed in every respect in the fabrication and assembly of the ETU reactor, The final reactor for the ART will differ from the ETU reactor only in modifications brought about by the fabrication, installation and operation of the ART. Only those changes considered to be absolutely essential to the successful operation of the ART are expected to be made. SPECIFIC TEST OBJECTIVES The most valuable information to be obtained from the ETU will probably be the disclosure of unanticipated difficulties, such as interferences in assembly or installation or difficulties arising from misoperation of certain elements of the system under peculiar operating conditions, but several prime objectives have been estabdlished for the ETU test program. The first and most important will be a determination of the tendency of parts to warp, shrink, or otherwise distort during the original welding and assembly processes and during testing. Dimensional checks will be made on all parts during inspection prior to and during assembly, and the actual dimensions will be recorded. The most important sets of diménsions, from the standpoint of satisfactory reactor operation, are 75 those that offect the moving parts, particularly the pump impellers, Distortion might cause inter- ference between the impelier and the stationary elements of the pump, which, in turn, might cause malfunctioning of the pump. Dimensional data on the cooling annuti for the core, reflector, and pressure shells wiil also be recorded, |t is important that these annuli be held to close tolerances, both in local regions and for the complete assembly. Some types of deviation will have little effect, while other types could have quite serious effects. [t is not possible to state explicitly which ones of the many possible combinations of devictions from drawing telerances will be acceptable and which will not, but it is important to determine the genercl magnitude and direction of the distortions in the ETU resulting from assembly operations and from testing so that the importance of such distortions in the ART can be reascnably ap- praised. Unfortunately, the temperature distri- bution and hence the distorticn pattern in the ART during high-power operation wifl be different from those in the ETU, but the distortion during zero- power operation and during mest of the fow- and medium-power operation of the ART should be the same as that in the ETU. Further, many porticularly bad off-design and fransient conditiens can be simulated in the ETU, and the effects on distortion witl be studied, Flow tests on various components will be carried out during the assembly of the recctor, These tests, which moy be made with either water or air, are partly for checking the calculated pressure drops through the complex circuits and partly for calibrating the systems so that they will serve as ftiewmeters for work during the high-temperature testing. WARMUP AND SHAKEDOWN TESTING During the initiai warmup and shakedown testing of the ETU a considerable amount of test dota will be cbtained for use in evaluating the design. Datc will be taken on the pressure drops through important elements of the system and, most es- pecially, on the over-all pressure drop ot each of a series of given flow rates, and the results will be checked against the predicted values. Close attention will be given to the behavier of the liquid levels in the expansion tanks for the hot fluids so that the accuracy and the dependa- bility of the liguid-level indicators caon be de- 76 termined. The heat losses to various elements of the system will also be measured, and the effects of operation and position of the heat-barrier doors and louvers in the auxiliary cooling system will be determined. It is especially importont that the sensitivity of the heat losses near the zero- power condition be determined as a function of position of both the heat-barrier deors and the louvers. OPERATING TESTS Operation of the ETU should follow the program prepared for the ART insofar as possible; in particular, the pump speeds and hence the system pressures should be programed in the same way as is planned for the ART. Thus the initial operation will be carried out af low pump speeds, the NaK pumps being cperated at probably one-half speed and the fuel and sodium pumps at abeut 10% speed. After completion of the low- and intermediote-power simulation in the ETU, the pump speeds wiil be increased to full design operating values. During this simulated operation it will be possible fo obtain further test data on the fiow characteristics of the various systems, the heat balance data, and some indication as to the performance of the heat exchangers, par- ticularly those in the reflector-moderator cooling circuit. The precision with which the heat balance data can be obtained will be determined by checking the heat balance detc for the air, the NaK, and the sodium systems against each other. During the shakedown operations it will probably be desirable to determine the effect that cutting out one or more pumps will have on the liguid levels in the expansicn tanks. Also, it may be possible to conduct tests on the performance of the xenon-removal system, A carefuily programed series of thermal-strain- cycling tests will be included in the ETU oper- ational tests, Delineation of this program will be delayed until the stress analysis work is essentially completed so that significant tests can be made. Hence the precise temperature levels, pressure levels, and flow rates that should be used in this program connct be specified until around December 1954. REACTOR ASSEMBLY The reactor is made up of five major subase semblies: the reflector-moderater, the main heat exchanger, the north head, the isiand and south - e e e L e pressure-shell liner assembly, and the pressure shell. Each of these major secticns is to be assembled and then fitted together in the proper sequence fo produce a complete assembly, Many of the individual components of the subas- semblies present difficult fabricational problems because of their geometric shape and the di- mensional tolerances specified. A particularly difficult problem will be encountered in helding close tclerances on complicated weldments and on the thin-walled concentric shells which form the fluid passages and separate the components of the reactor. The most meticulous and rigorous inspection techniques available are to be used on materiatls and welds. Assembly of the reacter will start with the reflector-moderator, which is made up of the outer nconel core shell, the beryllium hemispheres, the strutering structure, the B,C layer, and the inconel reflecter shells., The upper and lower halves of the Inconel outer core shell will be welded together at the equator, and the upper coliar will be welded to the top of the shell. The Inconel spacers will be fitted on the inside surface of the beryllium hemispheres, and the beryllium will be fitted around the Inconel shell, The spacers on the outer surface of the beryllivm will then be installed, together with the canned copper- B,C patches, at the north end. The strut-ring assembly and the Inconel reflector shell that houses the beryliium wili then be welded together and to the outer core shell. Next, the B,C tiles wiil be positioned on the outer surface of the assembly and covered with the 1/1{ -in.=thick Inconel shell that serves as the boron jacket. The B4C tiles are to be placed in sheet metal containers, one half of which will be spot-welded to the surface of the shells and the other half will be slipped into the attached helf fo form a container around the tile. This operation will complete the reflector-moderator assembly. Special movable fixtures will be used to place the 12 tube bundles of the main heat exchanger arcund the reflector assembly., The units must be fitted into place simultanecusly and held in posifion s¢ that the north-head assembly may be lowered over the reflector-moderator—heat ex- changer assembly. The north head is o compli- cated weldment containing the fuel pump volutes and housings, the fuel-expansion tank, the core entrance header, the sodium pump volutes, and the seodium-to-NaK heat exchangers. The north head will be built up from subweldments of the pump velutes and header passages on two decks. Welding accessibility and welding seguerice to prevent excessive warpage of critical surfaces are the most difficult problems envisioned for this assembly at the present fime. Weldability medels that illustrate the steps involved in assembling the north head are shown in Figs. 35 through 41. The island and south pressure-shell liner as- sembly will be assembled by fitting the upper and lower beryllium sections together and placing the spacers on the beryllium surface. The upper and lower sections of the inner core shell will then be placed around the beryllium, and the equatoricl weld will be made. The upper island and the expansicn joint will be welded to the core shell to form the istand assembly, The south pressure- shell liner assembly will then be assembled with the shells containing the neutron shieldirg and welded to the island assembly. The islond will be inserted through the moderator assembly so that the bellows assembly will slip into positicn in the north head and sc that the southern pressure- shell liner will seat against the northern section at the equater. The equaterial weld will then be made, The upper half of the pressure shell will be lowered over the reactor assembly, and the lower pressure shell, with the laminated filter plates in place, will be brought into position. The girth weld will be made for joining the two halves of the pressure shell. The upper island connection will be welded to the upper pressure shell, and the heat exchanger header pipes will be welded to the pressure shell sleeves. The sodium expansion tank will be welded to the upper pressure shell, and the control rod sleeve will be welded at the top of the expansion tank to complete the reactor assembly. The assembly will include the lead shield in order to obtain a test of the support structure ond the cooling systems, The water shield will be omitted on the ETU fo ease procurement and installation problems. The many shield pene- trations for instrumentation, helium, off-gas, and other connections to the reactor and reactor shell appurfenances make the detailed shieid design and instatlation very difficult, 77 Fig. 35. North-Head Weldability Mode! Showing Lower Deck and Peripheral Ring, Step 1. Na -~ TO - NeK o HEAT EXCHANGER BUNDLE Fig. 36. North-Head Weldebility Medel, Step 2. 78 g b apeweT PHOTO 25543 “’!l?-x_.i_ . f‘“ TOP DECK OVER Ng~TO~NaK EAT EXCHANGER Fig, 38, North-Head Weldability Model, Step 4. 79 80 | FUEL PUMP VOLUTES F FUEL EXPANSION TANK WALL TOP DECK OVER FUEL VOLUTE Fig. 40. North-Head Weldability Model, Step 6. P PHOTO 25539 Ihc puve) [ PO | PrioTO 25542 PROTO 25540 Fig. 41. North-Head Weldability Model, Showing Another View of Step 6. REACTOR DISASSEMBLY Dimensional data will be taken at frequent intervals throughout the disassembly of the ETU. The first step after removal of the reactor from the test stand will be the determination of the key dimensions of the pressure shell. The over-all heights from the level of the pump mounting flanges to the north head and to the south head will be measured. A cut will then be made at the equator through both the pressure shell and the pressure-shell liner. A cut will also be taken through the thermal sleeves around the heat ex changer header outlet tubes at the south end, and the sleeve that attaches the island assembly to the north head will be severed. This will permit the removal of the island and the southern half of the pressure shell. The core shells, pressure shell, and the heat exchanger can then be ex- amined. Dimensional data on the pump wells, on the pump impellers, and on the areas in the vicinity of close impeller clearances will be taken. The welds that attach the pressure shell dome to the north head around the roots of the fuel pump barrels will be milled out, arnd the transfer tubes that attach the reflector assembly to the north head, together with the thermal sleeves and heat exchanger outlet tubes at the north head, will be cut. The twe regions can then be separated for inspection. The reflector shell will be cut, and the beryllium will be removed and inspected for corrosion, spacer fretting, and thermal cracking, Other elements of the ETU system, including the snow traps, the celd traps, the filters in the NaK system, efc., wiil also be inspected. Typical sections of piping will be removed and inspected for corrosion and mass fransfer. The NaK pumps will be checked for changes in impeller running clearance, the dump valves will be inspected, and all elements of the plumbing will be carefully dye-checked for thermal cracks that might have been induced during the cperation. 81 Part V CONSTRUCTION AND OPERATION CONSTRUCTION AND OPERATIOR PLANS FOR INSTALLATION OF THE ART The present plans for assembly and installation of the ART were evolved to minimize the time re- quired to get the ART intc operation. It was recognized that not only must the ART assembly start before completion of the ETU tests but also that operation of the ART would have to be de- bayed if operation or disassembly of the ETU indicated the need for modifications of the reactor unit. As a result, three complete reactor units are being ordered. Subcssemblies of the third unit will be storted as soon os the second unit is assembied, but fina! assembly will not be started. The second reactor will be assembled and installed in Building 7503 as rapidly as possible and, if no trouble develops in the ETU, will be ploced in operation as soon as the disassembly and the inspection of the ETU have been completed. How- ever, if disassembly of the ETU indicates the necessity for modifications, work will be started en the third set of subassemblies as scon as decisiens are reached as to the nature of the modi- fications required. Final assembly of the third unit would proceed concurrently with the removal of the secend unit. OPERATION OF THE ART Filling eand Heating The initial fiiling and heating of the reacter are part of both the testing procedure and the operating procedure by virtue of the reactor being of the circulating-fuel type. The NaK system will be filled with NaK, ond, then, with the heat-barrier doors closed, the NoK pumps will be started. The power provided by pumping the NaK at one-half design-point flow will supply most of the heat re- quired, and electrical heaters will supply the rest of the heat needed to bring the system up to 1200°F in @ minimum of 24 hr, The sodium for cooling the reflector-moderator and the sodium for cocling the control rod will be added during the heating peried when the system temperature is about 350°F. With the sodium pumps operating, it will then be permissible to add more heat to the system through the main NaK system. With the system isothermal at 1200°F, the cold- trap systems will be gradually cooled down in order fo remove oxides from the NaK, The fused-salt fuel carrier will be put into the fill-and-drain tank when the entire system is in an isothermal condition ot 1200°F. The fuel pump will be started and will be operated at a nominal speed of 100 rpm, ond the fuel carrier will be pressurized info the reactor and heat exchanger system, The dump valves will then be closed, and the fuel and sodium pump speeds will be ine creased in order to degas the systems, When the system has been degassed and the design fuel has been obtained in the swirl tank, the main NaK pump speeds will be increased to design point. will be made stepwise in about six steps from one- half speed to design-point speed. level Increases in the main NaK pump speeds Isothermal operation with all pumps operaring at design-point speed will be maintained for about 24 hr. With all the pumps operating at design point the added power will raise the temperature of the system. The electrical heating will be decreased, as required, in order fo maintain the system in an isothermal condition. If it is necessary, a main blower will be started and the heat removal will be controlled by manipulation of the auxiliary louver positions so that the system can be maintcined in an isothermal condition with all pump speeds at the values desired, After no more than 24 hr of isothermal operation of the system, the dump valves will be opened and the fuel carrier will be dumped into the fili-and- drain tank. Opening the dump valves will autc- matically reduce the main NaK pump speeds to one- half design=point speed and at the same time will reduce the fuel and sedium pump speeds tc estab- lished minimums, ' Samples of the fuel carrier, the NaK, and the sodium will then be taken and examined for corro- sion products. Mass spectrographic analyses of these samples will also be made in order to de- termine whether any leakage occurred between any two adjacent systems, [f there is no evidence of leakage and if the corrosion-product analysis indicates that fthe carrier is clean, a portion of the carrier will be withdrawn from the fiil-and-drain tank to moke room for the addition, in two steps, of sufficient N02UF to provide a fuel mixture which hos about 40% of the U235 required to achieve criticality with the reguloting rod withdrawn., The estimate of 60% is 85 based on the system volume and on the data ac- quired during the high-temperature critical experi- ments described in Part H of this report. Enriching to Critical After addition of the initial charge of Na UF,, mixing will be accomplished by pressurizing the fluid into the system until its surface level is approximately at the mid-point of the reactor; the pressure will then be released 1o cllow the fluid to return to the fill-and<] NORMALLY OPEX o] NORMALLY CLOSED FUEL RECOVERY TANKS {1250"F WHEN CHARG - ING FIRST BATCH NITROGEN SUPPLY- SEE DIAGRAM 10 . PRE-POWER FUEL FILL-AND-DRAIN SYSTEM FUEL RECOVERY SYSTEM J i SAMPLING SYSTEM ‘ l FUEL ENRICHING SYSTEM J ‘ AND REACTOR ‘ l POST-POWER SAMPLING SYSTEM ‘ Flow Diagram 1. Fuel Fill-and-Drain, Enriching, Sampling, and Recovery Systems, 109 —r e a2 et T ™ R e e — e WATER RETURN {SEE DIAGRAM NO. 121 Y- HV“WN 23 wY HOCV X Wh-21{ & SET POINTS Hee l I | | I I | . ! I I | l—neuuu sum.v JSEE DIAGAAM NC.9) ..........._._..,.,,_.___._..._.___.____I v EOUIPMENT LpCATED QUTSIDE CELL | T P {2+min. TRANSIT TIME BETWEEN‘ SODIU"“‘ AND AUXILIARY EQUIPMENT PIT} it S v SET POINTS: L $ 20 (P& ] - 210 WATER SUPPLY ( SEE DIAGRAM NC. 12) TO {ELL *T“ /Hl ~FILLED HQUSING FOR CONTROL ROD DRIVE MECHANISM “WLSET AT 25 psig TO CELL TO TEMPORARY VACUUM PUMP ‘ ! " : w-rafiy 20 & T CELL VENT LINE (SEE DIAGRAM NG, 3) E : senomu:l 517 w32 OPENS AT 50 paig (SET POINTS) (SET POI TS5} Ne EXPANSION TANX (AUXILIARY) EATED LINES o~ nwrua: DisKs HELIUM SUPPLY (SEE DIAGRAM uo 51 AND VENT LINE TEMPORARY HELIUM SUPPLY N-21 HVAGN-t4 2 TEMPORARY vAGUI| if'r:luc ,._,J.___ " - KEATED LINE HELIUM SUPPLY (SEE DIAGRAM NO.2} @—— — —-Yon-31 HVXGN-!Z VN-3 LITHIUM FILL LINE No FILL LINE (HEATED) "~ W w GN-12 (; > /mr LINE FOR REMOVING No ' Nt mfiiuuc SUPPLY (SEZ DIAGRAM NO. 8) — i v o g o NYDRAULIC SUPPLY {SEE DIAGRAM lo. 4. (SEE O1AGRAM KO, 51 TO MYDRAULIC PUMP DISCHARGE PRESSURE — (SEE DIAGRAM ND. 4] Oll. SCAVENGE LINE LUBE OIL RETURNg BREATHER - L -5 @ @ uAmTE’f"é CRuP L LUSE QIL SUPPLY Ne EXPANSION TANK 3 mgnetc For — @— _Eié 200-600 (STANDBY, 300 °F ; OPERATE, 550 °F TOTAL VOLUME, 60 in? ; DIFFERENCE W VOLUME BETWEEN UPPER PROBE LEVEL AND LOWER PROBE LEVEL, 40 i} ) g3 ek ORML- LR-DWG 18080 EQUIPMENT IDENYIFICATION €S| CONTINUOUS SPEED INDICATOR CSR CONTINUOUS SPEED RECORDER CV CHECK VALVE ECV ELECTRICALLY ACTUATED CONTROL VALVE LI LEVEL LIGHT HCY HAND ACTUATED CONTROL VALVE HY MANUALLY OPERATED BLOCK VALVE NPA SODIUM PUMP A NPB SODIUM PUMP B NXA SODIUM-TO-Nek HEAT EXCHANGER A KXB SODIUM-TO~NoK HEAT EXCHANGER B Pl PRESSURE INDICATOR SV SOLENOID VALVE PT PRESSURE TRANSMITTER CFI CONTINUOUS FLOW INDICATOR CLI CONTINUOUS LEVEL INDICATOR PI CONTINUOUS PRESSURE INDICATOR 4 PRESSURE ALARM | POSITION INDICATOR-~LIMIT LIGHTS RELIEF VALVE FLOW ELEMERT FLOW ALARM FLOW INDICATOR . TEMPERATURE AL ARM WATER FILTER HAND -ACTUATED CONTROLLER HAND ~ACTUATED SWITCH TEMPERATURE -ACTUATED CONTROLLER PY PRESSURE ~-ACTVATED BLOCK VALVE mM ANIO P m < o X 353 PRESSURE, prig TEMPERATURE, *F FLOW, 1it/24 hr FLOW, gpm HEATED LINE - PV N-11 %] | .| B 153 1250 440 1050 287 1250 440 1050 % AUXILIARY NaK SYSTEM {SEE DIAGRAM NO. 6) "B W MAIN NoK SYSTEMS {SEE DIAGRAM NC.6} 3 a7 NOTE! SODIUM FLOW RATES ARE THOSE ASSGCIATED WITH ONE SODIUM PUMP MAIN NaK SYSTEMS {SEE DIAGRAM NO. 6) CELL d T ' : DA % \i“ nz - ws) B 1 1250 No DRAIN LINE [HEATED} {SEE DIAGRAM NO.1} Flow Diagram 2. Sedium stg?m. AUXILIARY NoK SYSTEM (3EE DIAGRAM NO.6) MAIN NaK SYSTEMS (SEE DIAGRAM NO. 6) MAIN Nok SYSTEMS {SEE DIAGRAM NO.6) FUEL FILL AND DRAIN LINES o T e —————— it S TO HYDRAULIC PUMP DISCHARGE BRESSURE {SEE DIAGRAM NO. 3} 4= OIL SCAVENGE LiNE LUBE OIL RETURN BREATHER LUBE OIL SUPPLY (SEE DIAGRAM NO.S) = 100> 5 txg 0 LINE BREAK DEVICE BIMETAL TEMPERATURE ELEMENT THERMOMETER CAP WELDED OM LINE THERMCGCOUPLE SWAGELOK CAP ON TUBE NORMALLY CPEN VALVE NORMALLY CLOSED VALVE THROTTLING VALVE SOLENOID VALVE 1M [ T ———— e ——e - ———— e ————— e ———— i-'ss I < 1 PI T L i 332 'SEE PROCESS AlR FLOW DIAGRAM 8 : I ‘{x-ss S :x-ia ' ) ‘ wm .-} ACTIVITY HS ACTIMITY -..___.- --..- ACTIVITY . ) | Ll - - MONITOR MONITOR [ | (=103l [=07%chl. L _] MONITCR ' " l S x-ss X-57 {X—IG ‘ C e - | R @_ FILTER ‘ FILTER FROM REACTOR FUEL PUMPS OIL cn‘rcu aAsms PR _ ] - : o SEE DIAGRAM NO. 8~ 7 e . X-55 ‘ : o $xe1d L Co . DOWNSTREAM SAMPLE F— - . DOWNSTREAM SAMPLE , < e . L STATION —REACTOR - ‘ _ * STATION-REACTOR o ol e - VENT SYSTEM I " 1Y} LAZY ROD o VENT SYSTEM X-42 LAZY ROD ST T © o X-83/\ TCONTROL ‘ Hv /\ CONTROL N | o ‘ T x-54 ' . X-13 | -.'"“ ) | o ‘ | OISASTER iN CELL Ll ‘ o PRI AT S _TO AUTOMATICALLY o , o e I | o CLOSE BLOCK VALVE E‘@__ . o Y 1] e ST L . . o X S I l L UPS_;‘TREAM SAMPLE fmmmmmereerery ' - . : " STAMON=REACTOR . S : HC] . VENT SYSTEM ] ] ‘l L T xeeY v L : L 8 A E g g ~ ‘ T 'SEE FUEL SYSTEMS' FLOW. l l a TV L e o REACTOR VENT SYSTEM - . IAGRAM NO.V = ° " : | l 3| . 'chYx-gb: x-8 sev | b Tav 1 o LAZY R oy I L el CONTROL ‘ PE l ‘ ,_H A FROM REACTOR FUEL R | I EED ple PUMPS BREATHER LINES oy o - - SEE DIAGRAM NO. 5 : l T ™~ CELL VENT SYSTEM X-42 | ‘. Ry ‘ . " %-44 E ' 1. . sCV | T __DISASTER IN CELL ) 70 AUTOMATICALLY - . \ | CLOSE BLOCK VALVE [“_J CELL AMBIENT ————f™} ' ' < V! : Xv g W Hla ‘CELL x-asng—:l—@ [F} - pev _ X..J L1 SEE SODIUM SYSTEM - FLOW DIAGRAM NO. 2 - g' P13 70 | | | | | | | | I | | | | | | | UPSTREAM SAMPLE p—emee——— STATION ~CELL VENT SYSTEM ‘E"E - 1 X-47, X-49.4 HV X-48 LAZY ROD CONTROL VENT HOUSE Flow Diagrdm 3. Off-Gas System. e e T e e — ] | ] | I I I | | | | | ! ! I. | ! | I | ! | | | i | I | LY HV_ DRAIN X-51 ! b _u_i‘-‘.u o/ \/ ‘ _l r WX-3 —_— e CHARCPAL ADSORBER IN CHARCOAL ADSORBERS THE DASHED LINES REPRESENT EMPYTY PIFE AND HEAVY LINES REPRESENT CHARCOAY FILLFD PIPE. JO>MDMZ%XXX ACTIVITY MONITOR iperTT. ORNL-LR-DWG 1505! LEGEND SWING CHECK VALVE HAND-ACTUATED SWITCH - MULTIPOINT ACTIVITY RECORDER MULTIPOINT TEMPERATURE INDICATOR HAND-ACTUATED CONTROL HAND-ACTUATED BLOCK VALVE PRESSURE INDICATOR PRESSURE TRANSMITTER PRESSURE ~ACTUATED BLOCK VALVE RADIATION ALARM DISASTER~ACTUATED BLOCK VAWVE VALVE OPEN VALVE CLOSED THROTTLING VALVE SOLENOID VALVE CHECK VALVE ° EQUIPMENT IDENTIFICATION ° PRESSURE (psig)} . FLOW (ctm). TEMPERATURE , *F ELECTRICAL CIRCUIT - WATER SEE WATER SYSTEM DIAGRAM NO, {2 SEE. WATER SYSTEM DIAGRAM NO. {2 113 v ¢ FLUD PUMP SPEEC INDICATOR {SEE DIAGRAM 1} SNUBBER AT 7.8 peig 86,2 hp MaX 2400 cpm MAX PUMP HYDRAULIC MOTOR HMA-1{ TO FPA (SEE DIAGRAM £) XX TO SODIUN PUMP SPEED INDICATOR (SEE DIAGRAM 2] e ——— AL ‘ (AND \LEADS N JHIS BULKKEAD W S -1z . SNUBBER OPENS AT 7.8 psiq 30 hp MAX 3300 rpm MAX PUMP HYDRAULIC MOTOR HMA-2 TO NPA (SEE DIAGRAN 2} HY H-i8 HV H-26 "TANK DESIGNED FOR 210 peig MAX . . PRAGLIC FLUD" WORKING PRESSURE, 135 paig wyl FILL—AND-DRAIN LiNE HYDRAULKC FLUHD FILL~AND~-DRAIN LINE TO FLUID PUNP INDICATOR (SEE DIAGRAM 1) A ¢ AT 1758 paig 5000 - MAX SNUBBER ‘ x 1 o - 95.2 np NAX 2400 rpm MAX HYDRAULXC PUMP HYDM.I:UC MOTOR TO FPB {SEE DIAGRAM 1} MOTOR BREAKER TO SODIUM PUMP SPEED INGICATOR HY WH-3 . {SEE DIAGRAM 2} SNUBBER AT 7.5 psig 30 hp MAX 3300 rpm MAX HV H-45 PUNP HYDRAULIC MOTOR HME - TO NPg (SEE DIAGRAM 2} X = - u -~ o w £g 2 B3 & . = 2 g |== [74] 1 H g 3 5:3 - §9 a 2 W i TANK DESIGNED FOR ,é 8 fi §§ H-36 [ ®o | m 210 paig MAX 5 2 ° €8 WORKING PRESSURE, 135 paig - a HYDRAULIC FLUID HYDRAULIC FLUD 8 FILL-AND-DRAIN LINE FILL~ AND« DRAMN LINE a <88 Flow Diagrafin 4. Cell Pumps Hydraulic Drive Systems, HY - L Y — o - TA = *TY — N - 5 - S = 2 A O nowom £ UNCLASSIFIED ORNL-LR~0WG 13082 CONTINUOUS CURRENT INDICATOR FLOW ALARM FLOW INDICATOR LEVEL INDICATOR MULTIPOINT TEMPERATURE RECORDER HAND =ACTUATED BLOCK VALVE PRESSURE INDICATOR ’ PRESSURE TRANSMITTER HAND-ACTUATED CONTROL VALVE CONTINUOUS POSITION INDICATOR TENPERATURE ALARN TEMPERATURE - ACTUATED CONTROL VALVE WOBBLE-PLATE POSITION SYNCHRO POSITION SWITCH ELECTRICAL SWITCH POSITION INDICATOR TEMPERATURE SWITCH CONTROL STATION ) ADJUSTABLE POSITION SWITCH LEVEL TRANSMITTER SWING CHECK vauvE TEMPERATURE WONCATOR FLOW ELEMENT PRESSURE, peig PRESSURE, in. Hg TEMPERATURE, 5F @2 ~— THERWOCOUPLE M THERMOCOUPLE WELL ¥———- THERMOCOUPLE ~—{}— NORMALLY OPEN VALVE ——p—— NORMALLY CLOSED VALVE —G— THROTTLING WALVE - CHECK VALVE el PRESSURE-RELIEF VALVE e~ BMETAL TEMPERATURE ELEMENT et MAIN WYORAULIC FLLIO LINES ———— . ELECTMICAL CONNECTION €= = THERMOMETER N, N/ o 00 om-0eF LigeTs 115 ar CELL ORHL -LA - DWE 15083 OUTSIDE BUILDING AUXILIARY EQUIPMENT RCOM TUNNEL WV _OR-47 OIL RETURN LINE ! " »R . L : NV, L OR-37 A OIL RETURN LINE o ) _ -— HVy LOR-27 OIL RETURN LINE J 2 b . - ‘ ] g [ ) WY 4OR-17 OIL RETURN LINE - 4 J 2 WV V0. 42 ] 3 3 ke [ ] vo-a3 y0-44 : VENT TO CELL >4 < € - ') 1 vo-33 vo-34 HVa gv0-32 . ) . ] ] _ ‘ 18 o 9-3: VENT TO CELL . - €~ ) ‘J N J T3 ' ‘J vo-23 vo-24 : uv=‘vo-zz _ l 0-2 e . S ) I [ < T3 HV . VO-12 ] ~ ] % lves ;lo-u HELIUM BYSTEM T 1. o 1 | . v L e e | 2] ] o] 5 K w 0-12 . 0-10 ’ J ] } v MY | = ] @) @) ] CED T OR-16 NIK LUBE O Hv}[o..s. Hv 160-50 HY160-10 29 . ; FILL LINEISEE - Vo= . e GIAGRAN T} Hv S e \ : - Ot vO-50 vo-10 co e Reservoir | HY " | LEVEL INDICATOR . o ‘ SPARE Lus— QAMMA SKIELD l=—1 GAMMA SHIELD bee—t- GAMMA SHIELD [ GAMMA SHIELD . —— oo . : *- . *- *- ‘ L o T-5 e e ] ’ \ 2 . SUPPLY ' ' o =] ; a L o . . - b~ T 2 ‘ Nok LUBE GIL -, ToTaL OF 28 TOTAL OF 28 TOTAL OF 28 ToTAL OF 28 ToTAL OF 28 10) W — 3T HEUTRON *- Mo s NERTRON i {55901} _ DRAINLINE(SEE qat OF OIL gal OF OIL gol OF OIL g0l OF OIL DIAGRAM T} . ‘ Hv JOR-86 HvIoR-59 i v , - L - L . . € WATER IN {SEE DIAGRAM 12} )J g vo-90 WATER QUT (SEE DIAGRAM 12) HY : o r. | | ' ol on CATCH CATCH 7OR-92 1 TANK TANK A Hv | oc-4 oc-2 on OFF ¢ i :q g OR-30 OR-93 OR-T3 Wy HY 1 TO BUCT DOWNSTREAM @ . OF RADIATORS (SEE L vent ko VENT TO vo-62 DIAGRAM B} w-12 ) WOTOR CcEL CELL o : STARTER STARTER TYPICAL FOR MOTORS _ TG CELL VENT SYSTEM ‘ : ORAIN LINE . 3¢ 440v 6.5 amp AT i SUPPLY LINE '(':E"EE:CM'::A?:: ,G" SYSTEM {SEE DIAGRAM 3) WASTE OIL " 10 PORTABLE fULL LOAD I SUPPLY LINE i . SUME TANK . REMOVABLE . N ey A i i 105 EQUIPMENT IDENTIFICATION SET BY MANUFACTURER 1 g FOR MAXIMUM OPERATING CFI CONTINUCUS FLOW INDICATOR o FLOW, Iiters/24 br TEMPERATURE.{ALL MOTOR CLI CONTINUOUS LEVEL INDICATOR . 4 TEMPERATURE ALARMS) €S CONTROL STATION O FLOW, gpm v v v - GV CHECK VALVE Vi3 JHY HYY Me13 017 LHV o A A 4 FA FLOW ALARM (€80 % OESIGN FLOW C—"3 TEWPERSTURE, °F DESIGN FLOW WiTH . :oa-i?:onm) & PRESSURE, prig FOR PORTASLE TaNk~"| - FE FLOW ELEMENT X THERMGCOUPLE ' [ ra0 | HCV HAND-ACTUATED CONTROL VALVE l g -1-< fl | WV HAND-ACTUATED BLOCK VALVE ——-Z= BIMETAL STRIP POSSIBLE FUTURE CONNECTIONS v o] i T4 LEVEL INGICATOR s s : QO O on-oFf LionTs L] L1 L] + LT LEVEL TRANSMITTER MO OR-2% | | ‘ MTR MULTIPOINT TEMPERATURE RECORDER . ELECTRICAL CIRCUIT 1 = oF -2 P e pressure INDICATOR ] R TR ’ TA TEMPERATURE ALARM —Pf4— PRESSURE RELIEF VALVE [ ] ] [ - { TCV TEWPERATURE CONTROL VALVE D~ VALVE OPEN ~\ L1 T 5 SUPPLY LINE ! RFE RESTRICTED FLOW —-w U 1 FE I — —fi—— ' PCV PRESSURE CONTROL VALVE —N— VALVE CLOSED AND-ACTUATED SWITCH [ P ] FILTER HS HAND-ACTUATED SOLENOID VALVE L .- - } . LE LEVEL ELEMENT 1 OR-44 OR-45 4 i J\ SUPPLY LINE " FI FLOW INDICATOR —P<]~ THROTTLING VALVE l!j =ij oy e o | AFM ACCUMULATIVE FLOWMETER m Hev FILTER { FM FLOWMETER -~ !- SWING CHECK VALVE - NORMAL ‘ VALVE POSITION I Flow Diagram 5. Reactor Pumps Lube Oil System. 117 B R T Do e i ——— »85 Nok ACTIITY MONITOR JO-oul SYSTEM 30-gol SYSTEM MOTOR BREAKER MOTOR BREAKER MOTOR SREAKEN . TAWI?!R TACHOMETER' TACHOMETER GENERATOR GENERATOR oL CATCH i GENERATOR OIL CATCH - LT TANK e TANK O'van‘u;c" - M OCA~3 - - 0cB-3 3550 rpm . OCA-1 PRE-POWER PRE-POWER OIL. DRAIN ey OIL DRAIN A \ \ \ PRE-POWER CiL DRAIN ToR NoK ACTIVITY MONITOR NaK ACTIVITY HS MAX 8% MiIN FILTER RFE RADIATOR . {SET_POINT . £0.8) f FILTER Fll;"l;lt.!_3 ' KFA-1 5 18 13 12 gol) TG A SEE DIAGRAM T HELIUM SUPPLY PURGE GAS VENT B PUMPS TO INTERMITTENT) FiLL~ WATER AND COMPRESSED AIR RETURN TC ENTRAINMENT SEPARATOR —SEE DIAGRAM 12 WATER SUPPLY TO FILTERS — SEE DIAGRAM 12 COMPRESSED AIR SUPPLY TO PLUG INDICATORS AND COLD TRAPS —SEE DHAGRAM 1Y AUXILIARY SYSTEMS ’——l———. MAIN SYSTEMS A Flow Diagram 6. Main, Auxiliary, and Special NaK Systems, AGRORET CRNL~LR - DWS 130540 TO OR FROM REACTOR 119 r - S| 90 MAX 89 MiN 129 Max 3 MIN 121 MAX 93 MIN Ha MaAX L THIS CIRCUIT COOLS THE OUMP TANK JACKET (CASHED LINES DENCTE MATERIAL HEATE ° OTHER THAMN Nox) 70 gt SYSTEM . oty e e o S — . A W p— o — T — T — i — - s . . e BREAKER WT!‘H GENERATOR GENERATOR Ot CATEH TANK ocA-4 OIL DRAIN Mokt ACTIVITY NaK ACTiVITY RADIATOR BANK PLW NDICATOR KA -2 ™ MY ECONOMIZER HEB-4 i ————————— i ——— A1 ——— — FILTER KFA-4 FILTER KFB-4 NoK FILTER DRAM LINE WH-21 | i FILTER Sev DRAIN LINE - HY - FILLING EQUIPNENT VR-87 TEMPORARY FILLING EQUIPMENT o e e e e o e o e k. S i et S — o —— — e ekl TR ot ot i o i S 15 1+ (112 gal) MR TO PLUG WATER TO PLUG INDICATORS AND WATER RETURN TO DRAN ; MAIN SYSTEMS A SPECIAL SYSTENS (IN MoK COOLING CHWAMBER) —4—- Flow Diagram 6 (continved) P ORNL+LR=-DWG 15034p MAIN SYSTEWS 8 121 ¥ TO OR FAROM REACTOR TO gal SYSTEM MOTOR BREAKER Iléu‘ OMETER GENERATOR s ON. CATCH © qashe TANK 3550 rpm oce-2 : PRE-POWER =TT oL DRAIN Loy - NaK ACTIVITY MONITON i PLUG INDICATOR ' RIB-2 ECONOMIZER KEB-2 MoK FILYER "y ORANY LINE K-83 TEMPORARY FILLING EQUIPMENT s 143 12 gai) VIATIRE RETURN TO DRAIN COOLING WATER 10 PLUG INDICATORS AND FILTERS L) COMPRESSED AIR TO PLUG n{mcm)fls AND FILTERS |_-—— MAIN SYSTEMS 8 ] VK-80 HV Nok ACTIVITY MONITOR KLB-11(C) 70 gul SYSTEM (SET POINT <1100} TACHOME TER GENERATOR (SET POINT 20.8) e FILTER KFB-4 TEMPORARY FILLING EQUIPMENT 5l {112 goi} FILL AND Flow Diagram 6 (continued) VK- SCV TO pucY SEE DIAGRAM & ™ DUCT SEE DIAGRAM 8 obecTET ORNL-LR-DWG 15034¢ CFR -~ CONTINUOUS FLCW RECORDER €5 ~ CONTROL STATION SCV — SWING CHECK VALVE Fl = FLOW INDICATOR FE ~ FLOW ELEMENT LUl = LEVEL INDICATOR 21= POSITION INDICATOR MAN = MULTIPOINY ACTIVITY RECORDER AFE - RESTRICTOR FLOW ELEMENT HV ~ HAND-ACTUATED BLOCK VALYE Pl - PRESSURE INDICATOR CS! = CONTINUOUS SPEED INDICATOR AWM ~ RECORDING WATTMETER FA = FLOW ALARM LA - LEVEL ALARM LMD ~ LEVEL—MEASURING DEVICE CLA = CONTINUOUS LEVEL RECURDER PT ~ PRESSURE TRANSMITTER HE ~ HAND-ACTUATEQ CONTROL MFI—~ MULTIPOINT FLOW INDICATOR M3 = HANO~ACTUATED SWITCH HCV = HAND=ACTUATED CONTROL VALVE CLI - CONTINUOUS LEVEL INDICATOR EV ~ ELECTRICALLY ACTUATEC BLOCK VALVE GF — HELIUM FILTER ] remeeratune, °F —————w— NaK LINES, NAJOR FLOW —_— NaK LINES, FILLING, FILTER, WDICATOR, ETC. gt PREUMATIC LINE e ELECTRICAL CONNECTION PRESSURE, p8i9 SWAGELOK CAP ON END OF TUBE SOLENDID VALVE _NORMALLY OPEM VALVE CONTROL VALVE NORMALLY CLOSED VALVE FLOW, gpm FLOW, ctm AT GO0® F AND 30 in. Mg FLOW, liters /24 hr LINE REOUCER OR EXPANDER z +OOO:xxz~T[> CHECK VALVE 123 L A b —— »- v ok ACTIVITY ©MONITOR wi=21 | o evER scv DRAIN LINE 185 ¥ 12 qat) MAIN SYSTEMS A iz T T T -4 120 max WP 93 MIN ORNL-LA-DWG 150484p 12t MAX 23 MiIN {8 MaX WIN FROM PUMS PUMP WEATED {DASHED LINES DENOTE MATERMAL THIS CIRCUIT COOLS THE DUMP TANK JACKET OTHER THAN NoK! 35 Mk - { TO gal SYSTEM e L s s S S St S s St s i o s S — T k. g et o st e A e e L e i S s e A8 A e e i e e e ) M . | worow sneaxn } ‘ z MOTOR BREAKER TACHOMETER . GENERATOR e OIL_CATCH : OK. CATEH TANK TANK OCA-4 oce-4 PRE-POWER ! PRE-POWER O, DRAIN | HY 0Oil. DRAIN 25 n? (184 gol} @- ' AR TO PLUG INOICATORS AND FILTERS WATER TO PLUG INDICATORS AND FILTERS WATER RETURN TO DNAIN ! SPECIAL SYSTEMB {IN SPECIA| NeK COOLING CHAMBER) _-‘L—= MAIN SYSTEMS @ { Flow Diagram 6 (continued) ; ] 121 MAIN DUCT ——— e —— SPECIAL EQUIFMENT DUCT COMPRESSED AIR SUPPLY {SEE DIAGRAM NG 11) X et apipr———— o e e« ¢ oo e, l . ALL HEAT BARRIER DGORS HYDRAULICALLY ACTUATED MoTOR —_— - L = ALL LOVVERS uvom;uucnuv ACTURTED " ‘:E: :{?f am ':,‘:,‘,: & N A n S ; :Q & DUCT-ANNULUS COOLING AIR } ‘ Lo ——— 7 MT-ANP:I..?S’MI - MACKFLOW DAMPERS gfi_ Ny (3F7] fi g PDT 013 - ] eed TN i : [ 2 = T | R L S =F B ChX —Ts — 7 MOTOR 1= TTTTTTTIS & woton] I~ | : Yras . _q 230mp Q i . '& L] | ¥ MAN BLOWER Aaa—c. ! :I . oLs _.‘{ —e & . 1 * ; . ’ i1 S ® -E Y 82,500 t : > @' __fi =R ) il OR QF | 3 = - uo;’on I i é ¢ i - - Lo as;” P B ': : I @‘—2 = N e i WAIN BLOWERABA-2 { l } a __ P . . B f t 1 -:.: L] 2 , I | : | " . ) s @% [ gl 1 B3 . i 1. lozs | - : wavor | [~ T : fy1 I & A BN e g s — - S . _ MM BLOWEN A0B-( ! : ' :: "'——nnntonbzvl.:-nscnou N\ MEAT BARRIER 000N N~ maotarons “NC hEaT BarmiER posRS . . P \ - ‘ E—“‘J_ _ _ Lt - . 82,300 oo 1E =% N @ @ (5] Lo VENT l ! ' . % RADIATOR AIR BYPASS —— i I = o — iz e — . I : L N s [ - M, —— - WAk GLEWER ABb-2 Ej E&c‘:g::m::u\:: . . £ e TR " . . . a4 10,000 oo r S i % RADIATOR AR BYPASS — moron § ‘\\\ '." . — = ‘ bo-eem- L | £ad— "3,' Q — o === - N BUCT-ANNULUS COOLING AIR ' DUCT-ANNULUS BLOWER ABA-3 Tt = Flow Diogram 8, b o . ., ‘ i 8 AR SAMPLE LINES TO NoK LEAK ~Ja ]l fxal e i, . . e e i, e, il WA e DETECTOR AND ALARM o0 1T 4000 venr ¥ m o {70 BE DEFERMINED) ¢ l - .. — O G DUCT-ANNULUS COOLING AIR | o-—OPPOSED-ACYION LOUVERS: by -« - 12,800 4 RADIATOR AIR BYFASS (SEE IAGRAM NG. 8) NeK OUT NaK IN —p ) ———— e e e 2l [ _D: »_ MOTOR 1 - ] v —————] - : & "‘ison. —— | ] i - «ra-a ] TRl i — b 2|4 % LS IQ’ - 2 — [sPR] m E———— 9 E n’sm:m. BLOWER ABA-4 = o % @_ - 12,500 D: m | | ! _[sPr] ——— [6zs] - L] ] - LI =2 , Y9l s /= | v MTR I’ - i 1 e AT ‘ KRB-4 ——x : — - - — : éb MOTOR i w14 _ _[skR] _= 30 hp -—X | L] ] — SECIAL BLOWER ASB-4 L%— —————— L M MEAT BARRIER DOORS N RADIATORS "N HEAT BARRIER DOORS el —_—————— jolrm ( Cr— 430 4000 i o t:r < @ RADIATOR AIR BYPASS B o< ) ] TVENT recs B——Q | & ) K ’ - DUCT -ANNULUS COOLING AIR L wol I &2\ F —™ : : 5h IS / "1 it [~ rfim-muuws BLOWER ABA-5 Process Air System. ' MYDRAULIC CONTAINERS VENT (SEE DIAGRAM NO. 4} SO, OANL LA = OWE 18090 CELL OFF-GAS (SEE DIAGRAM NO.3) REACTOR OFF -GAS {SEE DIAGRAM NO.3) 2 O 3 o NeK PUMPS LUBRICAT REACTOR PUMPS LY ’ b NG SYSTEM RESERVDIRS (SEE DIAGRAM NO. 7) : TING SYSTEM SUMP TANK VENT (SEE DIAGRAM NO. 5] MoK DRAIN TANKS th‘f {SEE DIAGRAM NO. 6} NoK PUMPS YENT t‘J £ DIAGRAM NO. 8) WASTE LUBE OIL § MP YANK VENT (SEE DIAGRAM NO. 31 EQUIPMENT 1DENTIFICATION CLOSED-POSITION INDICATOR SWITCH CONTINUOUS PRESSURE -DIFFERENTIAL NECORDER CONTROL STATION CONTINUOUS TEMPERATURE RECORDER MULTIRCINT TEMPERATURE INDICATOR MULTIPOINT TEMPERATURE RECONDER OPEN-POSITION INDICATON SWITCH PHESSURE-DIFFERENTIAL TRANSMITTER PRESSURE INDICATON SHUTTER POSITION INDICATOR SHUTTER POSITION RECORDER SPRCIAL SWITEH (CENTRIFUGALY INTERMEDIATE - POSITION INDICATOR SWITCH CLOSED LIMIT SWITCH OPEN LiMiT SWITCH NeKt PRESENCE IN AtR ALARW NeK PRESENCE IN AIR INDICATOR PRESSURE, in. H0 FLOW, ¢tm TEMPERATURE, *F ON-OFF LIGHTS ADJACENT 70 ITEM INDICATED PANEL ~MOUNTED LIGHT TO INDICATE CLOSED SWITCH COMPRESSED AIR LINES THERMOCOUPLE THREE ~WAY VALVE SOLENOID VALVE 127 A e C SONPTITR— ORNL -LR-0OWG 15058 5100 MAX_ 1100 MIN 2 G-34 :5?\'; HV X 6-63 ‘ HV HV . ~$><} . TO REACTOR FUEL PUMPS ' ‘ ' ' - 6-60 PCV 6-62 (SEE DIAGRAM NO. 1) - ‘ . {wve-32 = E ' v Hv X6-67 . - EQUIPMENT IDENTIFICATION ! 631 T TO REACTOR SODIUMIPUMPS XD Hvxs-aa G-66 : (SEE DIAGRAM NO. 2 SCV SWING CHECK VALVE - [ ASV 1000~2000 : 2 FE FLOW ELEMENT A ' FI FLOW INDICATOR Y + - ' HY X G- 71 FM FLOWMETER . ‘ TO FUEL EXPANSION JANK LEVEL INDICATOR HS HAND-ACTUATED SWITCH 6-70 { SEE DIAGRAM NO.11 HV HAND~ACTUATED BLOCK VALVE Lo '+ 500-1000 g v PA PRESSURE ALARM HC HAND-ACTUATED CONTROL HY X G-75 A PCV PRESSURE CONTROL VALVE TO FUEL OVERFLOW LENE LEVEL INDICATOR POT PRESSURE~DIFFERENTIAL TRANSMITTER (SEE DIAGRAM NO.1)§ - - Pl PRESSURE INDICATOR TO FUEL RECOVERY $YSTEM PS PRESSURE SWITCH (SEE FUEL FLOW DIAGRAM NO.1) RFE 'RESTRICTOR FLOW ELEMENT ' ‘ ASY AUTOMATIC SHUTOFF VALVE TO POST POWER FUEL] SAMPLE SYSTEM (SEE FUEL FLOWDIA RAM NO.1) FLOW, liters /24 hr TC FUEL-ENRICHING SYSTEM AND FILL-AND-DRAIN TANK (SEE FUEL FLOW DIA ?RAM NO. 1} ) FLOW, cmf TO CONTROL ROD SObIUM SYSTEM {SEE DIAGRAM NO. 2 PRESSURE, psig TO LITHIUM INJECTIQN TANK (SEE SODIUM FLOW PIAGRAM NO. 2) OPEN VALVE CLOSED VALVE ® | [Pl o TO NaK FILL-AND-DRAIN TANKS ({SEE NaK FLOW PIAGRAM NO.6) 3 THROTTLING VALVE Y 4 ‘ CHECK VALVE 771 x DO | J; ' 10-12 REDUCER GR EXPANDER e HVXG-42 HV HY L G-22 v < ] —<} TO A NoK PUMPS HVX6-23 6-25 JHV G-39 \ PCV G-41 : {SEE NaK FLOW [BIAGRAM NO. 6) OPENS AT ~ 6-40 280 psig ‘ “ ' (P 2200 250 [ | A L 10-12 HV X 6-46 I P > O 8 NeK PUMPH {SEE NaK FLOWDIAGRAM NO. 6) I 1 — e 0 REACTOR PUMP LUBE OiL RESERVOIR U ReALITUN FU [ P1 | {SEE DIAGRAM NO. 5) l L I Hv X6-50 A 4 TO NoK LUBE OlL RESERVOIR G-47 " PCV G-49 {SEE NoK LUBE OIL SYSTEM DIAGRAM NO. 7) - VEHICULAR ENTRANCE : P G-48 AUXILIARY EQUIPMENT PIT Flow Diagram 9. Helium System, ‘ 129 st i et 8 s e e .C i 1 EXISTING RELIEF VALVE M-15"QcET POINT 1100 psig - EXISTING RUPTURE DISK;_ EXISTING EXISTING RELIEF VALVE; SET POINT 4100 psig EXISTING RUPTURE DISK; © mv - SET POINT 1125 psig ' HY i EXISTING M-22 . M-24 EXISTING EXISTING RELIEF VALVE; SET POINT HOO psig EXISTING EXISTING RUPTURE DISK; - . SET POINT 1125 psig HV HY b M-29 EXISTING M-27 . HV STATIONARY NITROGEN CYLINDERS, WILL HOLD 10,500 scf AT 1000 psig GAS TRAILER 30,000 scf AT 1800 psig GAS TRAILER, 30,000 scf AT 1800 psig EXISTING [ P | ] 4000 psig HY X M~-32 HY X M-42 i © e i S it 5 8 NOR 116 MAX 250 260 RELIEF PRESSURE 300 psig VEHICULAR ENTRANCE m . T} i M-43 4 7 :‘ HV JE-D o ( EV -, ‘DISASTER IN CELL .",TO AUTOMATICALLY " "CLOSE BLOCK VALVE HV e SPARE M-44 AUXILIARY EQUIPMENT ROOM INSTRUMENTATION BULKHEPDS (6) FUEL FIlL-AND - DRAIN UMATIC SUPPORT (SEE FUEL ISTEMS FLOW DIAGRAM NO.1) i ROGEN SUPPLY FOR L INSTRUMENTATION L’ROGEN SUPPLY FOR sv sv FILLING CELL L TO HYDRAULIC SYSTEM JUNCTION PANEL BULKHEADS (SEE DIAGRAM NO. 4) TC HYDRAULIC SYSTEM CONTAINERS (S5EE PUMP DRIVE HYDRAULIC SYSTEM PROCESS FLOW DIAGRAM NO. 4) ) TO MAIN, AUXILIARY, AND SPECIAL ‘NaK PUMP SUMP LEVEL INDICATORS .{SEE Naok SYSTEMS FLOW DIAGRAM NO.6) RADIATOR PIT >0 bt b4 CQRELGEME. ORNL~-LR~-DWG 15091 EQUIPMENT {DENTIFICATION HVY HAND~ACTUATED BLOCK VALVE Pl PRESSURE INDICATOR RV RELIEF VALVE PS PRESSURE SWITCH PA PRESSURE ALARM HC HAND-ACTUATED CONTROL XV DISASTER-ACTUATED BLOCK VALVE FE FLOW ELEMENT FI FLOW INDICATOR SV SOLENOID VALVE PDT PRESSURE-DIFFERENTIAL TRANSMITTER SCV SWING CHECK VALVE HMC. HAND CONTROLLER PCY PRESSURE -ACTUATED CONTROL VALVE HS HAND SWITCH AF FLOW RESTRICYOR FLOW, cfm PRESSURE, psig ~—-—= ELECTRICAL CIRCUIT PRESSURE -RELIEF VALVE SWING CHECK VALVE OPEN VALVE CLOSED VALVE SOLENOID VALVE' LINE REDUCER OR EXPANDER 131 e — e et et e e S e e e s Sy e T b ———— WATER RETURN {SEE DIAGRAM NO.12) 25 sctm Flow Diagram 11, Compressed Air System, UNCLASSIFIED ORNL-LR-DWG $5092 EQUIPMENT IDENTIFICATION HCV HAND-ACTUATED CONTROL VALVE HV HAND-ACTUATED 8SLOCK VALVE PC PRESSURE CONTROLLER' SCV SWING CHECK VALVE HC HAND~ACTUATED CONTROL VALVE Pl PRESSURE INDICATOR PA PRESSURE ALARM PCVY PRESSURE CONTROL VALVE PS PRESSURE SWITCH LCY LEVEL-ACTUATED CONTROL VALVE LC LEVEL CONTROLLER RF FLOW RESTRICTOR FLOW, scfm PRESSURE, psig VALVE OPEN NTATION - SHOP VALVE CLOSED THROTTLING VALVE CHECK VALVE FOUR-WAY VALVE - WATER SUPPLY {SEE DIAGRAM NO.12) .‘, <3 . c-12 S AIR [ P | ' CHEMISTRY LABORATORY - - 1 5 HV c-51 ! - o . 5. SOLID STATE'LABORATORY By b c-13 - CPA i AlR Hv HC LB COMPRESSOR K LIMIT TO - N : RF , 8 sctm ‘ S o c-14 c-15|F =16 HVAC-18 m uv¥c-s2 -~ YO PROCESS AIR BACKFLOW DAMPERS ‘ HV PCV THV (SEE DIAGRAM NO, 8) : c7 b “ HC -Hv¥wc-52 g 0 I C-19 c-20| C-2t SR Y pwe- - 4 2 TO MAIN, AUXILIARY AND SPECIAL NaK PUMP SUMP LEVEL INDICATORS h 53 4 HY Y (SHE FLOW DIAGRAM NO. 6) ———— c-22 e 7we- FEL e ol MIST o >4 scv . 54 ”]I I TR Ny HV : b TO AUTOMATICALLY CLOSE A HCV I we-51. y 3 S ON "8" POWER FAILURE fi HV XWC-50 % L . 2. .’.‘A“’..ng"\:;':':_‘:;m;:“ — e TO NaK PLYG INDICATORS AND COLD TRAPS i ey 2 200 (SEE NoK BYSTEMS FLOW DIAGRAM NO. 6) i ] 20-30 e | ——— b Ev :j<: iy o} i c-10 c-t c-24 | ¥ . PEW POINT INSTRUMENTATION - AUXILIARY EQUIPMENT PIT A c-65 | ~20°F) ' ~ - r 11 [P ] . ! [ | c-61 ! INSTRUME NTATION - AUXILIARY EQUIPMENT PANEL PCV PS A - AIR cP8 ' [ ] 15-20 AIR l COMPRESSOR o REACTIVATION - c-26 : ! AIR SUPPLY INSTRUMENTATION - AMPLIFIER RACK fl HvXCc-~62 t HV { -l : 15-20 v H nv¥we-60 K il | INSTRUME 133 PR —— e " - - — s . ————— s A et R et e B L e e —— T Ap——— e s, T e et e e —— e e —_ et e T 1 ( 1 § N . HCV X W-80 HCV ) W-B2 Hc[v W74 HOY AW-76 wev Xw-g6 | e HH FE FE FE FE o | HY - ’ j -] . - © | ey . : WATER SHIELD A W OVERFLOW & ‘ ToPT LEAD SHIELD Y —pi] LINE - (W34 W3S s - 7 T /4 : N/ 7 7 . % . 1 o . w-ao¥uv- w-stYHv w-53XHV wW-55XHv : EXTERIOR o INSTALLATIONS s o - “ F w-63 1 ! { WHEN SWITGH } o ¥ CLOSED). . l 7 HY ¢ o ") ot . [ .l‘_" wn [y ) . ZA P P E gl= o0 ! l > Chi zx3 geyy S 7 e abg 4333 To Nk coLD TRAPS— | Q;E 7 EF’E Eéfié SEE DIAGRAM & | . X 2 a4 e ’ < o34 e ef5a L | | ; p S ' L | AP ~gme FIRE HYDRANT ‘ - | Y ' . . ) ! X "‘ L [/, ' 2 e fiee HOSE CONNECTION . . HeY, FE 2 v ——— ‘ . . : w-92 { BOTTOM) £ s AIR CONDITIONER . ‘ ; s e 1 r _ LUINCH ROOM ; TO STORM DRAIN i @- | HY e e EXISTING AIR COMPRESSOR - 10 CONTROL ROD Y \ W-87 ; ' o L R gomum VAPOR CON~ Y L ‘ MTR i ENSER —SEE prmrar————e - of HOT WATER HEATER .“ DIAGRAM 2 . . i . - R . TO REACTOR VENT HY [ 1| R~CONCH TIONING COMPRESSO v © LINE WATER JACKET ‘ l . ! SEE DIAGRAM 1 w-17 : ; l 1 C { | AND LAVATOR ‘ . . pmsror—gme- TOILETS AND L. ES TOR%"'TAELSEWESR ; ‘ " ' THROUGH AIR GAP! _ . e SHOWER o 7O REACTOR SNOW TRAP — ' S TO FILL AND DRAIN ——8h— SEE DIAGRAM 1 7o v TAléKDSk‘BD;JTR"AP— V ¥ MTR . ' o SEE DIAGRAM A N A | ‘ , L_9 - SERVIGE SINK = : J} " AIR CONDITIONER S — \ ‘ ‘@J__——fia—q- ok SOLID STATE LABORATORY CLOSED) [ TO STORM DRAIN [ za | E T HY L 4 NOT 1N USE oo h w-3 _ ‘ | = CHEMISTRY LABORATORY - - " e e o W-50XHV w-52 X Hv W—54 X HV f2s] W] b [WR] _ .__ W] — SAFETY SHOWER | L L [ L 1 2 W=7 ’ ] g | W-93 %PV W97 \HY W-83 X HY w91 w85 X HV w--8+ X HY Al 4 HY ' 1 HY HY \ HY et b [ — - I' w-64 xv w-gs| ! L W-85 Lo__|MmaR | ! ‘ - I [ ] HY X W-95 ;. F— Lo . T HV F | - = v X w-24 . wy W-67 MONITOR | - FUEL FILL AND WY L NOTE: DISASTER N CELL - . DRAN TANK w-58 o TO AUTOMATICALLY CLOSE STRAINER l e T BLOCK YALVES. . I~ LEAD SHIELD w-3 Y ‘ ‘ Scv / W-25 | h | TO STORM i -_| DRAIN : el Tcv ! uy . r FROM X~10 TO CELL 2 o i i RESERVOIR ANNULUS e TUNNEL : ‘ AIR L \ 4 AR ‘ 1 ) IN R , ' \ 0 | Hv XwW—ig ‘ | - = YO STORM DRAIN : i \’ TO CHARCOAL ADSORBER _ /' : OR HOLDING POND W-23 AND STORM ORAIN ~ o SEE DIAGRAM 3 . t‘ b | R ! . .[ !\ . \ PENTHOUSE ‘ ‘ VENT HOUSE ‘ k AUXILIARY EQUIPMENT AREA ‘ l CELL ‘ S Flow Diagram 12, Process Water System. NN ORNL~LR-DWG 15093 w-98 TO CELL DRAIN LSET POINT 100 psig) RUPTURE DISK LEGEND —m—————— SCV SWING CHECK VALVE £V ELECTRICALLY ACTUATED BLOCK VALVE FE FLOW ELEMENT - Fl FLOW INDICATOR HCY MAND-ACTUATED CONTROL VALVE HY HAND-ACTUATED BLOCK VALVE MT} MULTIPOINT TEMPERATURE INDICATOR MTR MULTIPOINT TEMPERATURE RECORDER eA PRESSURE ALARM Pl PRESSURE INDICATOR ZA POSITION ALARM s POSITION SWITCH LCV LEVEL-ACTUATED CONTROL VALVE MAR MULTIPCINT ACTIVITY RECORDER PX PRESSURE SWITCH TCV TEMPERATURE-ACTUATED CONTROL VALVE TE TEMPERATURE ELEMENT XV DISASTER-ACTUATED CONTROL, VALVE FLOW, gpm EQUMPMENT IDENTIFICATION ==~ THERMOCOUPLE IN WELL TEMPERATURE , *F PRESSURE, psig @DU ORIFICE (& TEMPERATURE-SENSING BULB " REDUCER NORMALLY OPEN VALVE THROTTLING VALVE NORMALLY CLOSED VALVE FIXXY RELIEF VALVE 135 ———— | —em ¢ r——— Appendix B PHOTOGRAPHS OF REACTOR MODELS AND STAGES IN THE CONSTRUCTION OF THE ART FACILITY ........ HECRET | PHOTOC 23368 i B-1. Preliminary Model of ART Recctor, Heat Exchanger, and Pressure Shell Assembly. 139 B-2. Secticn Through the Core, Reflector, and Island of the ART Model. 140 171 F . oo INSTRUMENT BULKHEAD SUPPORT| COLUMN | FLOOR | GRATING B-3. Model of ART, Including NaK Piping and Cell. | ~srcREE [ § PHOTO 25608 8 142 EACTOR § R SHIELD ATER \u*' NSTRUMENT HA ] S MEBER C — - O G o = 9y N 5 LU co ith Shield Cut Away to Show Reactor. Model of ART w . B-4 evl B-5. Plastic Model of ART North Head. ~SEEREP PHOTO 26473 7l B-6. Plastic Model of Nerth Head Showing L.ead Mocked Up in Wood. SRR PHOTO 26474 | : "UNCLASSIFIED o o " PHOTO 15410 B-7. View looking north showing the rear wall of the original Building 7503 and the early stages of excavation for the extension tc accommodate the ART. In the foreground is the deep excavation for the reattor cell. Behind it can be seen the excavation for the instrument and control tunnel. (Confidential with caption) 145 ori UNCLASSIFIED PHOTD 1581 B-8, View locking south from the original Building 7503 showing early steges of excavation and placement of concrete forms. On the right can be seen the forms for the charcoal adsorber tank. The columns and beam on the left will support the low-bay portion of the building extension. UNCLASSIFIED PHOTO 15694 1] : : ¥ =] ey b e ™. -1 . ¥ o B-9. Same as in Fig. B-7, 43 days later. In the center can be seen the reinforcing steel for the reactor cell support pad; behind is the form for the instrument and control tunnel. To the right can be seen forms for the specirometer room and the columns and beams to support @ low-bay extension. In Lyt the bottom center is shown the reinforcing steel extending up from the charceal adscrber tank, {(Confidential with caption) 8Pl 3 UNCL ASSIFIED PHOT O 15920 B-10, View looking south showing the reinforcing concrete footings for the stack and work in progress on placement of support columns for the high- bay extension. In the right center is the charcoa! adsorber tank, and in the axtreme left center is the form work for the spectrometer tunnel, e SO & | ENCLASSIFIED PHOTO 15758 B.11. View looking north showing the completed reactor cell support pad and the instrument and control tunnel. At the top center is shown construce 6¥1 tion in progress on the switchhouse. The form work for the top haif of the charceal adsorber tank is shown in the lower left. (Confidential with caption) 0sl UNCLASSIFIED PHOTO 16060 B-12. View looking south toward the stack foundation. In the foreground is shown support columns for the high bay and form work for the radiator pit wall. To the right of the columns can be seen excavation for the blower house. - UNCLASSIFIED PHOTO 16058 B-13. View locking scutheast showing the spectrometer room at the lower left and the spectrometer tunnel in the center, both of which are loceted under the low-bay extension to the building. 151 Zsi UNCLASSIFIED PHCTOC 16169 B.14. View looking north showing the steel framing for the building extension., In the fereground can be seen the stack base and the top of the char- coal adsorber tank. The switchhouse can be seen at the left. eat B-15. View looking northeast showing the early construction the main air duct for the primary cooling system. UNCLASSIFIED PHOTD 16280 & & : 5 work on the blower house. The opening in the main building at the right will recaive CLASSIFIED 070 16317 K B' 16- View lcoking north taken from the south end of the high-bay extension and showing an initial stage in the eraction cof the reacior cell water tank. The curved stee! ponels centairing nozzles, in the upper left, will be joined into a ring and located on the anchor bolts shown ot the bottom center and thus become the support ring for the inner vessel of the reactor call. {Confidentia! with caption) 154 ICLASSIFIED ] PHOTO 16456 § B-17. View locking southeast showing three of the five spectrometer tubes which provide a beam path between the reactor cell and the spectrometer tunnel, which is underground beneath the lew bay. {Confidential with caption) 155 951 PHOTO 17152 B.18. View looking southeast across the construction work on the reactor cell, The form work, right center, is for the special equipment room walls. The fiil shown in the top center is over the spectrometer tunnel and spectrometer tubes. (Confidential with caption) LASSIFIED L UMC PHOTO 17213 B-19. View looking southeast across the reactor cell showing a step in the placement of the top hemispherical head on the inner vessel of the cell. The outer tank shown is the 30-ft-dia water tank. (Confidential with caption) 157 gs1 0 176 B-20. View looking north showing the facility in a late stage of the building alteration and addition program. In the right center is the 10sft-dic 75-fi-high stack for discharge of the air to be supplied by blowers to be located in the blower house at the bottom center. At the leftis thediesel- generator house, which will accommeodote the five auxiliary units required for ART operotion. LRI L= UNCL ASSIFIED w- PHETO 18233 B-21. View taken inside the blower house showing the supply end of the main cooling air duct and installation work on two of the four 82,000-cfm [ ) 1 .t o I £ =1 . 9 - ol 1T AAM £ st E e 1 01 1 . L N1 4 1 o -1 ' o8 1 . 2 .1 . DIOWeEerS.e vn LOTN 3 I108s Or TRe main qucy are me jv,vuv=crm pDIowers, WRICN Wil SUppIly QIr imro me GNnulius petwesn e Dulia g cencrerg ana e in= sulated steel duct for the purpose of preventing overheating of the building structure. At the lower left can be seen the opening for the ramp which leads 661 down to the radiator pit beneath the ceoling air duct. o9t B-22, View looking west and from inside the cooling air duct showing the stainless steel facing over the insulated steel liner of the duct, The supply end of the duct is the opening shown in the center of the photograph. {91 will 8‘230 is the View taken NCL ASSIFIED Q% 18153 inside of the main air duct from the base of the disch rge stack sh wing the porti n of the duct in which the NaK-to-air radiators At the right ¢ nter is the ¢t facing of the cell water tank and the through the NaK pipe w ] Hass the duct the tis the pit. Zol UNCL ASSIFIED PHOTO 17733 B-24. View looking south across the reactor cell. In the foreground can be seen the top of the cell inner tank. At the top right is the concrete pen - house which will house the primary coolant pumps and their drive motors. To the left of the penthouse are the roof plugs which will cover the special equipment room immediately below. {(Confidential with caption) £91 UNCL ASSIFIED PHOTO 17734 B-25, View showing the southwest corner of the auxiliary equipment room. This roem, which was used during ARE operation as the heat exchanger pit, is being modified to serve as the auxiliary equipment room and will house the lubricating cil pumping system and the hydraulic drive equipment, y9l n STy 2 iy 2 e = gl Pt TUNCL ASSIFIED PHOTO 17966 ottt it IR M Pyttt B-26. This view, which was tcken in the switchhouse, shows the primary switchgear and instruments associated with the receipt and distribution of the two power supplies to the facility. The switchgear and instruments on the left will serve the incoming purchased power from TVA, Those on the right are for the locally generated power. 991 L UUNCLASSIFIED . PHOTO 17960 B-27. View looking south from the north end of the original 7503 Building showing the top of the cell water tank in place as it will be located during ART operation. This photograph was taken during the vacuum testing operation on the inner tank. {Confidential with caption) 291 UNCLASSIFIE . PHOTO 17961 g B-28. View looking north across the cell while the water tank top is in place. At the lower left can be seen the penthouse structure and the area in which the main coolant pumps and motors will be located. (Confidential with caption) L91 UNCLASSIFIED PHOTO 17951 » ¥ B-29, View e opened cell. The rectangular plates on the floor structure of the inner vessel will support the reactor. The circular nlates will cunport the chislded fusl fillcandadrain tank ond fuslaracovery tank. The 24sin.edin nozzlae throush the side woll of the veczal will provide for the penetration of the instrumentation and control lines, lube oii lines, hydraulic drive lines, cooling water lines to the cell equipment. (Confidential with caption) , gas lines, and heater leads 891 B-30. View looking north across the completed and painted cell. The floor structure was removed from the inner vessel at the time of the photograph and therefore a good view is presented of the fluid distribution weirs located in the bottom of the vessel. The scalloped ring cutside the weirs is the support for the floor structure. At the extreme bottom of the photograph can be seen the NaK piping sleeves with their expansion joints which penetrate the two cell tanks. To the right of these sleeves can be seen portions of three of the 24-in.~dia spectrometer tubes. (Confidential with caption) UNCLASSIFIED PHOTO 18032 B-31, View locking southwest toward the modified bhuilding. In the foreground are the two top heads for the cell tanks which have been removed to 691 storage ouiside the facility. {Confidential with coption) 0LL UNCLASSIFIED 'PHOTO 18033 Be32. View looking southwest toeward the diesel-generator house. In the lower left is the substation for the 13.8-kv purchased power supply from the TVA system. The gas cylinders in the jower center of the photograph will be used for storage of nitrogen during ART operation. tZ1 B-33, View lcoking west iary power system. inside UNCLASSIFIED- 0 18154 the diesel-generator house showing a stage of the instaliation work on the five diesel-generator units for the auxil- BIBLIOGRAPHY BIBLIOGRAPHY (Publications listed chronologically by subject) GENERAL DESIGN Nuclear-Powered Flight, Lex-P-1 (Sept. 30, 1948). C. B. Eilis, The Atomic-Powered Aircraft January, 1950, ORNL-684 {April 26, 1950}. Report of the Technical Advisory Board, ANP-52 (Aug. 4, 1950). A. P. Fraas, Effects of Major Parameters on the Performance of Turbojet Engines, ANP-57 (Jan. 24, 1951). C. B. Ellis, The Technical Proh;’gms of Aircraft Reactors, ANP-64, Part | (June 8, 1951), Part 1 {Jan. 2, 1952}, G. F. Wislicenus, Hya’roa’ynamzcs of Homogeneous Reactors, ORNL CF-52-4-191 (April 30, 1952}, W. B. Cottrell (ed.), Reactor Program of the Aircraft Nuclear Propulsion Project, ORNL- 1234 {(June 2, 1952), A. P. Fraas, Three Reactor Heat Exchanger—Shield Arrangements for Use with Fused Fluoride Circulating Fuel, ORNL Y-F15-10 {June 30, 1952). i~ A. P, Fraos and M. E. LaVerne, Heat Exchanger Design Charts, ORNL-1330 (Dec. 7, 1952). A. P. Fraas, Compenent Tests Recommended to Provide Sound Bases for Design of Fluoride Fuel Reactors for Tactical Aircraft, ORNL. CF-53-1-148 (Jan. 13, 1953). W. S. Farmer, Minimum Weight Analysis for an Air Radiator, ORNL CF-53-1-111 (Jan. 31, 1953). R. C. Briant, Objective and Status of ORNL-ANP Program Presentation USAF Advisory Committee, Washington, January 12, 1953, ORNL CF-53-2-126 (Feb. 13, 1953). A. P. Fraas and C. B. Mills, A Reflector-Moderated ("zrculatz:zg +Fuel Reactor for an Aircraft Power Plant, ORNL CF-53-3-210 (March 27, 1953}, . W. S. Farmer et al., Preliminary Design and Performance Studies of Sodium-to-Air Radiators, 0RNL-]509 (Aug. 26, 1953). B. Lubarsky and B. L. Greenstreet, Thermodynamic and Heat Transfer Analyses of the Aircraft Reactor Experiment, ORNL-1535 (Aug. 27, 1953). R. C. Brian, Proposed ANP Program, ORNL CF-53-12-15 (Dec. 3, 1953). R. W. Bussard ez al., The Moa’emtzng Coolmg System for the Reflector-Moderated Reactor, ORNL-1517 (Jan 22, 1954). W. H. Jordan, Proposed ANP Program — Ada’endum, ORNL CF-54-3-119 (March 24, 1954). A. P. Fraas and B. M. Wilner, Effects of Reacror Deszgn Conditions on Aircraft Gross Weight, ORNL CF-54-2-185 (May 21, 1954). [ F. A, Field, Temperature Gradient and Thermal Stresses in Heat-Generating Bodies, ORNL CF-54-5-196 (May 21, 1954). A. H. Fox et al., A Reactor Design Parameter Study, ORNL CF-54-6-201 (June 25, 1954). A. P, )Frcas and A. W. Savolainen, ORNL Aircraft Nuclear Power Plant Designs, ORNL- 1721 (Dec. 3, 1954), W. B. Cottrell et al., Aircraft Reactor Test Hazards Summary Report, ORNL-1835 (Jan. 19, 1955). H. C. Hopkins, Vertical Components of Fuel Forces on Reflector and Island, ORNL CF-55-2-142 (March 2, 1955). C. S. Burinette, Fission-Product Heating in the Off-Gas System of the ART, ORNL CF-55-3-191 (March 28, 1955). A. S. Thompson, Empirical Correlation for Fatigue Stresses, ORNL CF-55-4-34 (April 5, 1955). 175 176 V. J. Kelleghan, High-Temperature Valve Information Summary, ORNL CF-55-4-83 (April 5, 1955), A. S. Thompson, Allowable Gperating Conditions, ORNL CF-55-4-44 (April 11, 1955). A. S. Thompson, Flexible Mounting Systems, ORNL CF-55-4-124 (April 11, 1955), E. S. Bettis, ART Design Data, ORNL CF-55-4-87 {April 18, 1955). W. B. Cottrell, The ART Gff-Gas System, ORNL CF-55-4-116 (April 21, 1955). A. S. Thompson, Thermal Stresses in Tube-Header Joints, ORNL CF-55-4-15% {April 25, 1955). T. J. Balles, Surface-Volume Ratios for Five Different Fluoride Fuel Systems, ORNL CF-55-5-93 {May 12, 1955). T. J. Balles, Temperatures and Pressures Resulting from Chemical Reaction Within the ART Reactor Cell, ORNL CF-55-5-135 {May 17, 1955). R. |. Gray end M. M. Ycorosh, Heat Transfer Design Data for ART Intermediate Heat Exchanger, ORNL CF-55-5-120 {May 20, 1955). R. . Gray and M, M. Yarosh, ART Main Heat Exchanger Designs, ORNL CF-55-5-141 (May 23, 1955). B. L. Greenstreet and A. $. Thompson, Permissible Rate of Temperature Rise of Shells, ORNL CF-55-5-175 (May 26, 1955). A. S. Thompson, Flexible Support of Massive Bodies, ORNL CF-55-7-101 (July 19, 1955). L. B. Andersen, Adsarption Holdup of Radioactive Krypton and Xenon, ORNL CF- 55-8-103 (Aug. 16, 1955). H. C. Hopkins, Estimate of Nuclear Energy Generated During ART Accident, ORNL CF-55-8-185 {Aug. 24, 1955), L. E. Anderson, Symmetrically Loaded Cylindrical Skell with Fixed Ends, ORNL CF- 55-9-37 (Sept. 7, 1955). J. M. Eastmen, Nuclear Aircraft Powerplant Controf, ORNL CF-55-9-121 (Sept. 23, 1955). R. L.)MaxweH, Analysis of Various Designs for Intermediate Heat Exchanger Header, ORNL CF-55-10-16 {Oct. 3, 1955}, W. J. Price, Radiation Levels Directly from Off-Gas Stack, ORNL CF-55-10-74 (Oct. 18, 1955). W. J. Price, Neutron and Gamma Flux Distributions in ART, ORNL CF-55-11-7 (Oct. 18, 1955]). M. H.)Cooper, Pressure Drop of Heat Exchanger Tube Spacers, ORNL CF-55-11-180 (Neov, 28, 1955), C. W. Dollins, Lead Shielding for ETU/ART, ORNL CF-55-12-102 (Dec. 19, 1955). W. J. Price, Preliminary Estimate of Activation of Various ART Materials by Thermal Neutrons, ORNL CF-56-1-191 (Jan. 12, 1956). W. L. Scott, Jr., Dimensional Data for ART, ORNL CF-56-1-186 (March 13, 1956). D. L. Platus, Thermal Stress Analysis for a Proposed South End Configuration in the ART, ORNL CF-56-5-166 {May 25, 1956}, F. A. Field, Rectangular Cross Section Cantilever Beams in the Plastic Range, CRNL CF-56-6-85 (June 20, 1956). COMPONENT DESIGN AND TESTING G. H. Cohen, A. P, Fraas, ond M. E. LeVerne, Heat Transfer and Fressure Loss in Tube Bundles for High Performance Heat Exchangers and Fuel Elements, ORNL-1215 (Aug. 12, 1952), R. E. Ball, Investigation of the Fluid Flow Pattern in a Model of the *'Fireball’* Re- actor, ORNL Y-F15-11 (Sept. 4, 1952}. H. R. Jchnson, Valve and Pump Packings for High Temperature Fluoride Mixtures, ORNL Y-F17-28 (Sept. 15, 1952). W. B. Cottrell, Components of Fluoride Systems, ORNL CF-53-1-276 (Jan. 27, 1953). W. B. Cottrell and L. A. Mann, Sodium Plumbking, ORNL CF-53-8-49 {Aug. 14, 1953). H. J. Stumpf, Design Data and Proposed Test Schedule for Sodium-to-Air Radiators, ORNL CF-53-9-102 (Sept. 8, 1953). R. W. Bussard, Flat Plate Heat Exchangers for Reactor System Use, ORNL CF-53- 10-208 (Oct. 26, 1953). B. M. Wilner and H. J. Stumpf, Intermediate Heat Exchanger Test Results, ORNL CF-54-1-155 (Jan. 29, 1954). H. J. Stumpf, R. E. MacPherson, and J. G. Gallagher, Test Resuits and Design Com- parisons for Liguid Metal-to-Air Radiators, ORNL CF-54-7-187 (July 19, 1954). J. E. Ahern, High-Conductivity Fin Fest Results, ORNL CF-54-8-200 (Aug. 27, 1954). E. S. Farris, Summary of High Temperature, Liquid Metal, Fused Salt Pump Development Work in the ORNL-ANP Project for the Period July 1950—]anuary 1954, ORNL CF- 54-8-234 (Aug. 1954). R. W. Bussard and R. E. MacPherson, Thermal Stresses in Beryllium — Test No, 1, ORNL CF-54-10-106 (Oct. 25, 1954), D. F. Salmon, Turbulent Heat Transfer from a Molten Fluoride Salt Mixture to Sodium Potassium Alloy in a Double-Tube Heat Exchanger, ORNL-1716 (Nov. 1954). R. |. Gray, Temperature-Time History and Tube Stress Study of the Intermediate Heat Exchanger Test, ORNL CF-54-11-69 (Nov. 30, 1954). L. A. Mann, ART Reactor Accidents Hazard Tests, ORNL CF-55-2-100 (Feb. 11, 1955). T. J. Balles, Coupling Between Probes Used for Flow Studies on the Metal and Plastic Core Medels for the ART Reactor, ORNL CF-55-11-117 (Nov. 18, 1955). J. C. Amos, Small Fluoride-NaK Heat Exchanger Test No. 1 (High Velocity Heat Ex- changer Test), ORNL CF-56-1-187 (Jan. 2, 1956}. J. M. Eastman, NaK Systems for Varying Reactor Load Coupling, ORNL CF-56-2-101 (Feb. 21, 1956). C. B. Thompson, Steady-State Control Characteristics of Chemical-Nuclear Aircraft Power Plants, ORNL-1976 {Feb. 29, 1956). J. J. Milich, ART Fuel Dump Valve Test, ORNL CF-56-4-42 {April 6, 1956}, A. L. Southern, Closed-Loop Level Indicator for Corresive Liquids Operating at High Temperatures, ORNL-2093 (May 17, 1956). PHYSICS C. B. Mills, Heating in the B ,C Curtain Due to Neutron Absorption and the B1%(n,a)Li’ Reaction, ORNL Y-F10-64 (zug 16, 1951). C. B. Mills, A Simple Criticality Relation for Be-Moderated Intermediate Reactors, ORNL Y-F10-93 (March 10, 1952). W. K. Ergen, Physics Considerations of Circulating-Fuel Reactors, ORNL Y-F10-98 (April 16, 1952). F. G. Prohammer, Note on the Linear Kinetics of the ANP Circulating-Fuel Reactor, ORNL Y-F10-99 (April 22, 1952). W. K. Ergen, The Kinetics of the Circulating-Fuel Nuclear Reactor, ORNLL CF-53-3-231 {March 30, 1953). W. K. Ergen, The Inbour Formula for a Circulating-Fuel Nuclear Reactor with Slug Flow, ORNL CF-53-12-108 (Dec. 22, 1953}. W. K. Ergen, The Bebhavior of Certain Functions Related to the Inbour Formula of Circulating-Fuel Reactors, ORNL CF-54-1-1 (Jan. 15, 1954). J. L. Meem, The Xenon Problem in the ART, ORNL CF-54-5-1 (May 3, 1954). - W. K. Ergen, The Kinetics of the Circulating-Fuel Reactor, ORNL CF-54-4~6 {May 5, S 1954). | 177 178 J. E. Fautkner, Calculation of Fission Neutron Age in NaZrF_, ORNL CF-54-8-97 (Aug. 31, 1954). 4. E. Faulkner, Age Measurements in LiFF, ORNL CF-54-8-98 (Aug. 31, 1954). P. H. Pitkanen, On Gamma Ray Heating in the Reflector-Moderated Reactor, ORNL CF-54-2-111 (Sept. 14, 1954). J. W. Noaks, Preliminary FEvaluation of Possible Poisons for Use in the ART Control Rod, ORNL CF-55-2-16 {Feb. 2, 1955). L. T. Anderson, Gamma and Neutron Heating of the ART Fuel Pump Assembly, ORNL CF-55-3-161 (March 28, 1955). L. T. Anderson, Calculation of the Beryllium Contribution to the ART Temperature Coefficient of Reactivity, ORNL CF-55-5-76 (May 11, 1955). W. K. Ergen, Probable Effect of Replacing Inconel by Columbium in ART Core Shells, ORNL CF-55-5-100 (May 13, 1955). A. M. Perry, The Boron Laver of the ART, ORNL CF-55-8-38 (Aug. 4, 1955). W. K. Ergen, Rate of Molecular Dispersion of Heat in the ART and of Dye in an Aqueous Solution, ORNL CF-55-9-40 (Sept. 9, 1855). H. W. Bertini, An Estimate of the Self-Abscrption of the Decay Gammas in the ART Fuel Dump Tank, ORNL CF-55-12-47 (Dec. §, 1955). H. W. Bertini, Activity of the Na Cooclant in the ART, ORNL CF-55-12-78 (Dec. 16, 1955). A. M. Perry, Fission Power Distribution in the ART, ORNL CF-56-1-172 (Jan. 25, 1956). W, K. Ergen, Uranium Investment in a Circulating Fluoride Power-Producing Reactor, ORNL CF-56-4-29 (April 4, 1956). A. M. Perry, ART Pile Period During Fill Operation, ORNL CF-56-4-34 (April 9, 1956). C. N. Copenhaver, Basic Inelastic Neutron Data for Age Calculations for Various Fuel and Moderator Compositions, ORNL CF-56-6-40 (June 6, 1956). A. M. Perry, Effect of Gaps in the Boron Layer, CRNL CF-56-6-65 {June 13, 1956). A. M. Perry, Burnup of Boron in ART and in Sample lrradiations, ORNL CF-56-6-153 (June 29, 1956). H. W. Bertini et al., Basic Gamma-Ray Data for ART Heat Deposition Calculations, ORNL-2113 (July 5, 1956). 51’ CRITICAL EXPERIMENTS D. Scott and B. L. Greensireet, Reflector Moderated Critical Assembly, Experimental Program, ORNL CF-54-4-53 (April §, 1954), D. Scott, The First Assembly of the Small Two Region Reflector-Moderated Reactor, ORNL CF-54-7-15% (July 26, 1954}, D. Scott, The Second Assembly of the Small Two Region Reflector Moderated Reactor, ORNL CF-54-8-180 (Aug. 25, 1954). D. Scott and R. M. Spencer, The Second Assembly of the Small Two Region Reflector Moderated Reactor {Part 11}, ORNL CF-54-9-185 (Sept. 27, 1954). B. L. Greenstreet, Reflector Moderated Critical Assembly Experimental Program — Part H, ORNL CF-54-10-119 (Oct. 19, 1954). R. M. Spencer, The First Assembly of the Three Region Reflector Moderated Reactor, ORNL CF-54-11-33 (Nov. 5, 1954}. R. M. Spencer, Preliminary Critical Assemblies of the Reflector-Moderated Reactor, ORNL-1770 (Nov. 22, 1954). B. L. Greenstreet and R. M. Spencer, The Second Assembly of the Three Region Re- flector Moderated Reactor, ORNL CF-54-11-150 (Nov. 24, 1954). R. M. Spencer, Three Region Reflector Moderated Critical Assembly (Experimental Results), ORNL CF-54-12-189 (Dec. 28, 1954). R. M. Spencer, Three Region Reflector Moderated Critical Assembly with 146-1'720 Inconel Core Shells, ORNL CF-55-1-123 (Jan. 21, 1955), J. W. Noaks, Preliminary Evaluation of Possible Poisons for Use in the ART Control Rod, ORNL CF-55-2-16 (Feb. 2, 1955). R. M. Spencer, Three Region Reflector Moderated Critical Assembly with I,é-z'n.. Inconel Core Shells, ORNL CF-55-2-93 (Feb. 14, 1955). J. W. Noaks, Fvaluation of Reactivity Characteristics of Control Rods and Materials Potentially Suitable for Use in the ART. Part I, ORNL CF-55-4-84 {April 13, 1955). J. W. Noaks, Fvaluation of ART Control Rod Materials, Part Ill. The Effects of Neutron Irradiation on Some Rare Earth Samples, ORNL CF-55-4-137 (April 25, 1955), J. W. Noaks, Fvaluation of ART Control Rod Materials, Part IV. The Variation of Reactivity with Control Rod Diameter, ORNL CF-55-4-178 (April 29, 1955). J. W. Noaks, Evaluation of ART Control Rod Materials, Part V. Addendum to Parts I-1v, ORNL CF-55-5-147 (May 20, 1955). E. V. Sandin, Reflector Moderated Critical Assembly Experimental Program — Part I, ORNL CF-55-5-181 {May 27, 1955}, D. Scott, J. J. Lynn, and E. V. Sandin, Reflector Mocderated Critical Assembly with End Ducts — Experimental Program (Experimental Results), ORNL CF-55-6-94 {June 16, 1955). S. Snyder et al., Three Region Reflector Moderated Critical Assembly with End Ducts and Y -in. Inconel Core Shells. CA-21-1 Neutron Flux Measuremeni, CRNL CF- 55-10-142 (Oct. 28, 1955). D. Scott et al., Three Region Reflector Moderated Critical Assembly with End Ducts. Experimental Results with CA-22, Enlarged End Duct Modification, ORNL CF-56-1-96 (Jan. 30, 1956). D. Scott et al., Three Region Reflector Moderated Critical Assembly with End Ducts — Experimental Results with CA-23, Enlarged Island Modification, ORNL CF-56-1-97 (Jan. 30, 1956). CORROSION L. S. Richardson, D. C. Vreeiand, and W. D. Manly, Corrosion by Molten Fluorides, ORNL-1491 (March 17, 1953). G. M. Adamson, Examiration of Bi-Fluid Loop No. I, ORNL CF-53-7-199 (July 29, 1953). H. Inolye, Scaling of Columbium in Air, ORNL-1565 (Sept. 1, 1953). G. M. Adamson and R. S, Crouse, Examination of LF Pump Loop, ORNL CF-53-10-117 (Oct. 6, 1953). G. M. Adamson and R, S. Crouse, Metallographic Examination of Second Heat Exchanger from Bi-Fluid Pump Loop, ORNL CF-53-10-228 (Oct. 27, 1953). G. M. Adamson, Metallographic Examination of Ferced Circulation Loop No. 2, ORNL CF-54-3-15 (March 3, 1954). W, D. Manly, High Flow Velocity and High Temperature Gradient Loops, ORNL CF- 54-3-193 (March 18, 1954). W. D. Manly, Interim Report on Static Liquid Metal Corrosion, ORNL-1674 (May 11, 1954). G. M.)Adamson and E. Long, Examination of Sodium, Beryllium, Inconel Pump Loops Numbers 1 and 2, ORNL CF-54-9-98 (Sept. 13, 1954), W. K. Stair, The Design of a Small Forced Circulation Corrosion Locp, ORNL CF- 54-10-97 (Oct. 11, 1954). G. M. Adamson and R. 3. Crouse, Examination of First Three Large Fluoride Pump Loops, ORNL CF-55-3-157 {March 24, 1955), G. M. Adamson, R. S. Crouse, and P. G. Smith, Examinatiorn of Inconel-Fluoride 30-D Pump Loop Number 4695-1, ORNL CF-55-3-179 {March 28, 1955}, 179 180 G. M. Adamson and R. S. Crouse, Examination of Sodium-Inconel Pump l.oop 4689-4, ORNL CF-55-4-167 (Aprii 21, 1955), G. M. Adamson and R. S. Crouse, Examination of Fluoride Pump Loops 4930-A and 4935-1, ORNL CF-55-4-181 (April 26, 1955). G. M. Adamson and R. S. Crouse, Examination of Fluoride Pump Loops Numbers 4695-2 and 4695-3, ORNL CF-55-5-97 (May 11, 1955}. R. S. Crouse, E. Long, and A. Toboada, Examination of Sodium Inconel Beryllium Loops 4667-3 and 4907-7, ORNL CF-55-5-123 (May 12, 1955}, G. M. Adamseon and R. S. Crouse, Examination of Inconel-316 Stainless Steel—Sodium Pump Loops 4689-5 and 4689-6, ORNL CF-55-6-24 (June 2, 1655). G. M. Adamson and R. S. Crouse, Effect of Heating Method on Fluoride Corrosion. Examination of Fluoride Pump Loops 4935-2, 4950-1, 4950-2, ORNL CF-55-6-99 (June 14, 1955). R. J. Gray, Evidences of Mass Transferred Material in Inconel and 316 Stainless Steel Sodium-to-Air Radiator, CRNL CF-55-6-89 (June 15, 1955). G. M. Adamson and R. S. Crouse, Examination of Short-Operating Time Loops Numbers 4695-4 and 4695-5, ORNL CF-55-7-66 (July 8, 1955), G. M. Adamson and R. §. Crouse, Examination of Loop 4950-3-Inconel with a UF and UF ; Mixture in Zirconium Fluoride, ORNL CF-55-8-70 (Aug. 9, 1955). G. M. Adamson and R. S. Crouse, Examination of Inconel-Fluoride Pump Loops 4935-3 and 4935-4, ORNL CF-55-8-94 (Aug. 12, 1955). G. M. Adamson and R. S. Crouse, Examination of Final Loops in Time Series Numbers 4695-4D(2) and 4695-5C(2;, ORNL CF-55-8-151 (Aug. 19, 1955). R. 5. Crouse, J. H. DeVan, and E. A. Kovacevich, Examination of Fluoride Pump Loops 4935-5 and 4935-7, ORNL CF-55-10-31 (Oct. 5, 1955). R. S. Crouse, J. H. DeVar, and E. A, Kovacevich, Examination of Miscellaneous Pump Loops 4950-4 and 4917, ORNL CF-55-10-32 (Oct. 5, 1955), R. J. Gray and P. Patriarca, Metallographic Examination of ORNL Radiator No. 1 and York Radiator No. 1 Failures, ORNL CF-55-10-129 (Oct. 31, 1955). H. lnouye, M. D’Amore, and J. H. Coobs, Boron Containing Materials — Inconel Com- patibility, ORNL CF-55-12-117 (Dec. 22, 1955). J. V. Cathcart and W. D. Manly, The Mass-Transfer Properties of Various Metals and Alloys in Liguid Lead, ORNL-2008 {Jan. 10, 1956). J. H. DeVan, Examination of SHF-C Resistance Heater, ORNL CF-55-5-145 (May 31, 1956). J. V. Catheart, L. L. Hall, end G. P. Smith, Oxidation Characteristics of the Alkali Metals; 1. Oxidation Rate of Sodium Between ~79 and 48°C, ORNL-2054 (May 22, 1956). R. J. Gray, Metallographic Examination of High Velocity Heat Fxchanger (SHE No. 1), ORNL CF-56-5-148 (May 24, 1956). T. Hikido, The Role of Chromium in the Mass Transfer Attack of Alloys by Fluorides, ORNL CF-56-6-58 (June 10, 1956). J. H. DeVan and R. 5. Crouse, Examination of Hastelloy B Pump Loops 7425-12 and -13, ORNL CF-56-6-79 {June 11, 1956). J. H. DeVan and R. S. Crouse, Examination of Pump Loops 7425-14 and 7425-15 Which Circulated Fuel 70, ORNL CF-86-6-161 (June 29, 1956). J. H. DeVen, Effect of Temperature on Depth of Attack in Inconel-Fuel 30 Pump Loops, ORNL CF-56-7-9 (July 5, 1956). W. D. Merly, Fundamentals of Liguid-Metal Corrosion, ORNL-2055 (July 12, 1956}, J. E. Van Cleve, Metallographic Examination of I.LH.E, No. 3, ORNL CF-554-7-82 (July 12, 1956). J. H. DeVan and R. S. Crouse, Examination of ORNL [ and 2 Intermediate Heat Ex- changers, Type IHE-3, ORNL CF-56-7-135 (July 20, 1956). J. E. Van Cleve, Metallographic Examination of §.H.E. No. 2 (ORNL No. }) Heat Ex- changer, ORNL CF-56-7-130 (July 27, 1956)}. CHEMISTRY T. N. McVay, Petrographic Examination of Fluoride Fuels, ORNL. CF-52-6-127 (June 20, 1952). C. J. Barten, Fused Salt Compositions, ORNL CF-53-1-129 (Jan. 15, 1953). C. J. Barton, Fused Salt Compositions, ORNLL CF-53-10-78 {Oct. 9, 1953), E. Orban, Data on the BeF ,-NaF System, ORNL CF-54-4-47 (April 27, 1954), R. F. Newton, Effects of Fission Froducts on Performance of a Reactor Using Fluorides as Solvent for Fuel, ORNL CF-54-5-40 (May 7, 1954). F. F. Blankenship, Analysis of Salt Mixtures, ORNL CF-54-5-189 (May 25, 1954). E. Orban, Measurement of the Stability of Lithium Hydride, ORNL CF-54-5-47 (May 286, 1954). C. J. Barton, Fused Salt Compositions, ORNL CF-54-6-6 (June 2, 1954), J. C. White, Sclubility of Composition 30 in Water, ORNL CF-55-4-18 {April 4, 1955), J. C. White, A. S. Meyer, Jr., and D. L. Manning, Differential Spectrophotometric De- termination of Beryllium, ORNL-1909 (June 24, 1955), J. C. White, Determination of Iron(lll) in Mixiures of Alkali Metal Fluoride Salt, ORNL CF-55-7-103 (July 22, 1955}. C. J. Barton, Fused Salt Compositions, ORNL CF-55-9-78 (Sept. 16, 1955), J. C. White, Determination of Traces of Iron(iIl) in NaFaLiF-KF-UF4 with Thiocyanate, ORNL CF-55-9-96 (Sept. 20, 1955). D. L. Manning, A. S. Meyer, Jr., and J. C. White, The Compleximetric Titration of Zirconium Based on the Use of Ferric Iron as ithe Titrant and Disodium-i, 2-Diby- droxybenze-3, 5-Disulfonate as the Indicator, ORNL-1950 {Sept. 28, 1955). W. J. Ross, A. S. Meyer, Jr., and J. C, White, Determination of Trivalent Uranium with Methylene Blue, ORNL-1984 (Nov. 7, 1955). J. C. White, Chemical Examination of Cold Trap from Intermediate Heat Exchanger Test Stund No. 1, ORNL CF-55-11-102 (Nev. 17, 1955). J. C. White, Determination of Small Amounts of Tantalum in NaF-LiF-KF and in NaF- LiF-KF-UF ., ORNL CF-56-1-49 {Jan. 10, 1956). J. R. Smolen, Physical Properties of Boral Used in Lid Tank Mockups, ORNL CF- 56-1-94 (Jan. 23, 1956). J. C. White ez al., Determination of Trivalent Uranium in Fluoride Sait Mixtures by the Modified Hydrogen Evolution Method, ORNI.-2043 (Feb. 10, 1956). J. C. White and A. S. Meyer, Determination of Traces of Aluminum in NaF-ZrF -UF ,, ORNL CF-56-3-10 (March 1, 1956). J. C. White, A. $. Meyer, and B. L. McDowell, Determination of Dissolved Oxygen in Lubricating Fluids, ORNL-2059 (March ¢, 1956). J. C. White, Procedure for the Determination of Oxygen in Sodium and NaK by the Distillation Metbod, ORNL CF-56-4-31 (April 5, 1956). P. A. Agron, B. S. Borie, Jr., and R. M. Steele, 83°1\‘T‘I2UF6’ ORNL CF-56-4-199 (April 30, 1956), R. E. Thoma, X-ray Diffraction Results, ORNL CF-56-6-25 {June 4, 1956}, J. C. White, Determination of Microgram Amounts of Titanium in the Presence of NaF- ZrF ;-UF ,, ORNL CF-56-6-111 (June 20, 1956). W. J. Ross, A, S. Meyer, Jr., and J. C. White, Determination of Boron in Fluoride Salts, ORNL-2135 (Aug. 7, 1956). 181 182 METALLURGY A. DeS. Brasunas, A Simplified Apparatus for Making Thermal Gradient Dynamic Cor- rosion Tests (Seesaw Tests), ORNL CF-52-3-123 (March 13, 1952). W. D. Manly, Status of Columbium, Beryllium, and Lithium, ORNL CF-54-4-162 (April 6, 1954). E. E. Hoftman and P. Patriarca, Heat Exchanger Fabrication, ORNL CF-54-5-88 (May 17, 1954). G. P. Smith, M. E. Steidlitz, and L. L. Hall, Flammability of Sodium Alloys at High Temperatures, ORNL-1769 (June 15, 1955). H. lnouye, Neutron Shieid for ART, ORNL CF-55-7-94 (July 18, 1955). R. J. Gray and P. Patriarca, Metallographic Examination of ORNL HCF Radiator No. | Failures, ORNL CF-55-10-129 (Oct. 31, 1955). H. inouye, Use of Boror for Neutron Skielding, ORNL CF-55-12-3 {Dec. 1, 1955). R. S. Crouse, Metallographic Examination of Thermal Convection Loops and Capsule Tests Performed by WADC, ORNL CF-56-2-26 (Feb. 6, 1956). J. H. DeVan, Preliminary Investigation of Inconel Castings, ORNL CF-56-2-56 (Feb. 13, 1956). R. J. Gray and P. Petriarca, Metallographic Examination of PWA HCF Radiator No. 2, ORNL CF-56-3-47 (March 12, 1956}, R. Carlander and E. E. Hoffman, Transfer of Carbon Between Dissimilar Metals in Contact with Molten Sodium, ORNL CF-56-4-73 (April 2, 1956). H. Inouye, M. D’Amore, and T. K. Roche, Nickel Base Alloys for High Temperature Service, ORNL CF-56-4-121 (April 16, 1956). M. R. D’Amore and H. lnouye, The Extrusion of Composite Tubes, ORNL CF-54-4-123 (April 18, 1956). R. J. Gray, Metallographic Examination of High Velocity Heat Exchanger (SHE No. 1), ORNL CF-56-5-148 (May 24, 1956). P. Patriarca et al., Fabrication of Heai Exchangers and Radiators for High Temperature Reactor Applications, ORNL-1955 (June 14, 1955), T. Hikide, Screening Evaluation of Experimental Nickel-Molybdenum Alloys, ORNL CF-56-7-35 (July 10, 1956). HEAT TRANSFER H. C. Claiborne, A Critical Review of the Literature on Pressure Drop in Noncircular Ducts and Annuli, ORNL-1248 (May 6, 1952), H. F. Poppendiek and L. D. Palmer, Forced-Convection Heat Transfer in Thermal Entrance Regions — Part II, ORNL-914 (May 26, 1952). W. S. Farmer, Cooling Hole Distribution for Reactor Reflectors, ORNL CF-52-9-201 (Sept. 3, 1952). W. S. Farmer, Heat Generation in a Slab for a Non-Uniform Source of Radiation, ORNL CF-52-9-202 (Sept. 4, 1952}, H. F. Poppendiek, Estimates of Heat and Momentum Transfer Characteristics of the Two Fluoride Coolants: (LiF-48 M %, BeFZ-_SZ M %) and (NaF-10 M %, KF-46 M %, ZrF ;-44 M %), ORNL CF-52-11-205 (Nov. 29, 1952). H. F. Poppendiek and L. D. Palmer, Forced-Convection Heat Transfer in Pipes with Volume Heat Sources, ORNL-1395 (Dec. 2, 1952). W. B. Harrison, J. O. Bradfute, and R. V. Bailey, Generalized Velocity Distribution for Turbulent Flow in Annuli, ORNL CF-52-12-124 (Dec. 19, 1952). W. B. Harrison, Forced Convection Heat Transfer in Thermal Entrance Regions, Part 11, ORNL-915 (Aug. 6, 1953). H. W. Hoffman, Preliminary Results of Flinak Heat Transfer, ORNL CF-53-8-106 {Aug. 18, 1953). H. F. Poppendiek and G. M. Winn, Some Preliminary Forced-Convection Heat Transfer Experiments in Pipes with Volume Heat Sources Within the Fluids, ORNL CF-54-2-1 {Feb. 1, 1954). J. O. Bradfute, The Measurement of Fluid Velocity by a Photographic Technique, ORNL CF-54-2-37 (Feb. 4, 1954), D. C. Hamilton, F. E. Lynch, and L. D. Palmer, The Nature of the Flow of Ordinary Fluids in a Thermal-Convection Harp, ORNL-1624 (Feb. 23, 1954), L. D. Palmer and G. M. Winn, A Feasibility Study of Flow Visualization Using a Phos- phorescent Particle Method, ORNL CF-54-4-205 {(April 30, 1954). H. F. Poppendiek and L. D. Palmer, Forced Convection Heat Transfer Between Parallel Plates and in Annuli with Volume Heat Sources Within the Fluids, ORNL-1701 (May 11, 1954). M. W. Rosenthal, H. F. Poppendiek, and M. R. Burnett, A Method for Evaluating the Heat Transfer Effectiveness of Reactor Coolants, ORNL CF-54-11-63 {Nov. 4, 1954). H. F. Poppendiek, A Laminar Forced-Convection Solution for Pipes Ducting Liguids Having Volume Heat Sources and lLarge Radial Differences in Viscosity, ORNL CF-54-11-37 (Nov. 5, 1954). D. C. Hamilton et al., Free Convection in Fluids Having a Volume Heat Source, ORNL- 1769 (Nov. 15, 1954), J. O. Bradfute, Qualitative Velocity Information Regarding the ART Core: Status Report No. 4, ORNL CF-54-12-110 (Dec. 14, 1954). H. W. Hoffman and J. Lones, Fused Salt Heat Transfer — Part ' Forced Convection Heat Transfer in Circular Tubes Containing NaF-KF-LiF Eutectic, ORNL-1777 (Feb. 1, 1955). H. W. Hoffman, i.. D. Palmer, and N, D. Greene, Electrical Heating and Flow in Tube Bends, ORNL CF-55-2-148 (Feb. 22, 1955). G. L. Muller and J. O. Bradfute, Qualitative Velocity Profiles with Rotation in 18-Inch ART Core, ORNL CF-55-3-15 (March 1, 1955). H. F. Poppendiek, Status Report on Forced Convection Experimental Work in Converging and Diverging Channels with Volume Heat Sources in the Fluids, ORNL CF.55-3-174 {(March 24, 1955). F. E. Lynch and J. O. Bradfute, Qualitative Velocity Profiles in 18-Inch ART Core with Increased Turbulence at Its Iniet, ORNL CF-55-5-132 (May 10, 1955). J. O. Bradfute, F. E. Lynch, and G. L. Muller, Fluid Velocity Measured in the ]8-Inch ART Core by a Particle-Photographic Technigue, ORNL CF-55-6-137 (June 21, 1955), F. E. Lynch, Qualitative Velocity Information Regarding the ]8+Inch ART Core with Vane Set No. 2 in Entrance, ORNL CF-55-6-173 (June 27, 1955}, G. L. Muller, Qualitative Velocity Information Regarding the 18-Inch ART Core with Turbulator Vane Set No. I at Entrance, ORNL CF-55-4-174 (June 27, 1955). L. D. Palmer, G. M. Winn, and H. F. Poppendiek, Investigation of Fluid Flow in Helical Bends of the ANP In-Pile Loop, ORNL CF-55-6-183 (June 27, 1955). G. L. Muller, Qualitative Estimates of the Velocity Profiles in the 21-Inch ART Core, ORNL CF-55-7-92 (July 20, 1955). D. C. Hamilton and F. E. Lynch, Free Convection Theory and Experiment in Fluids Having a Volume Heat Source, ORNL-1888 (Aug. 3, 1955). H. F. Poppendiek and L. D. Palmer, Application of Temperature Solutions for Forced Convection Systems with Volume Heat Sources to General Convection Problems, ORNL-1933 (Sept. 29, 1955). G. L. Muller and F. E. Lynch, Qualitative Velocity Information Regarding the Quarter Scale Model of the 18-Inch ART Core and the 5.22 Scale Model of the 2i-Inch ART Core with the P and W Swirl Nozzles at the Inlet, QORNL CF-55-10-48 (Oct. 3, 1955), 183 184 G. L. Muller and F. E. Lynch, Qualitative Velocity Information Regarding Two Constant Gap Core Models, ORNL CF-55-10-49 (Oct. 3, 1955). F. E. Lynch and G. L. Muller, Qualitative Velocity Profiles with a Rotational Com- ponent at the Inlet of the 2i-Inch Core Model, G. F. Wislicenus Design, ORNL CF- 55-10-50 (Cct. 3, 1955), 4. L. Wantland, Therma! Characteristics of the ART Fuel-to-NaK Heat Exchanger, ORNL. CF-55-12-120 (Dec. 22, 1955), J. L. Wantlond, A Method of Correlating Experimental Fluid Friction Data for Tube Bundles of Different Size and Tube Bundle to Shell Wall Spacing, ORNL. CF-56-4-162 (April 5, 1956). H. W. Hoffman, Thermal Structure for the Region Beyond the ART Reflector — Supple- ment 1, CRNL CF-56-4-129 (April 17, 1956). H. W. Hoffman, Fused Salt Heat Transfer, ORNL CF-56-7-85 (July 20, 1956). PHYSICAL PROPERTIES M. Tobias, S. |. Kaplan, and S. J. Claiberne, Densities of Certain Salt Mixtures at Room Temperature, ORNL CF-52-3-230 (March 26, 1952). J. M. Cisar, Densities of Fuel Mixtures Nos. 25 and 25a, ORNL CF-52-5-209 (May 28, 1952). R. F. Redmond and T. N. Jones, Viscosity of Fulinak, ORNL CF-52-4-148 (June 19, 1952). R. F.)Redmond and T. N. Jones, Viscosity of Fuel Sait Mixtures No. 27 and No. 30, ORNL. CF-52-7-138 (July 23, 1952). L. Cooper and S. J. Claiborne, Measurement of the Thermal Conductivity of Flinak, ORNL CF-52-8-163 (Aug. 26, 1952). W. B. Harrison, Transient Methods for Determining Thermal Conductivities of Liguids, ORNL CF-52-11-113 (Nov. 1, 1952), S. J. Claiborne, Measurement of the Thermal Conductivity of Fluoride Mixture No. 35, ORNL CF-52-11-72 (Nov. 11, 1952). W. D. Powers and G. C. Blalock, Heat Capacity of Fused Salt Mixture No. 21, ORNL CF-52-11-103 (Nov. 15, 1952). R. F. Redmond and T. N. Jones, Density Measurements of Fuel Salts Nes. 31 and 33, ORNL CF-52-11-105 {Nov. 15, 1952). R. F. Redmond and T. N. Jones, Viscesity Measurements of Fuel Salts Nos. 31 and 33, ORNL CF-52-11-106 {(Nov. 15, 1952). S. J. Claiborne, Measurement of the Thermal Conductivity of Fluoride Mixtures No. 30 and No. 14, ORNL CF-53-1-233 (Jon. 8, 1953). R. F. Redmond and S. |. Kaplan, Remarks on the Falling-Ball Viscometer, ORNL CF- 53-1-248 (Jan. 14, 1953). W. D. Powers and G. C. Blalock, Heat Capacity of Fused Salt Mixture No. 30, ORNL CF-53-2-56 (Feb. 6, 1953}, Pbysical Properties Charts for Some Reactor Fuels, Coolants, and Miscellaneous Materials, 3rd ed., ORNL CF-53-3-261 (March 20, 1953), W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 14, ORNL CF-53-5-113 {May 18, 1953). W. J. Sturm, R. J. Jones, and M. J. Feldman, The Stability of Several Fused Salt Systems Under Proton Bombardment, ORNL-1530 (June 19, 1953). S. i. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity of Fluoride Mixture No. 40, ORNL CF-53-7-125 (July 23, 1953). S. I. Cohen and T. N. Jones, Measurements of the Solid Densities of Flucride Mixture No. 30, Ber, and NaBeF ,, ORNL CF-53-7-126 (duly 23, 1953). W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 12, ORNL CF-53-7-200 (July 31, 1953). S. I. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity of Fluoride Mixture No. 44, ORNL CF-53-8-217 (Aug. 31, 1953). S. |. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity of Composition 43 (Na,UF ), ORNL CF-53-10-86 (Oct. 14, 1953). W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 33, ORNL CF-53-11-128 (Nov. 23, 1953). S. I. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity of NaFoZrF4=UF (62.5-12.5-25.0 mole %), ORNL CF-53-12-179 (Dec. 22, 1953). W. D. Powers anfi G. C. Blalock, Heat Capacity of Fuel Composition No. 31, ORNL CF-54-2-114 (Feb. 17, 1954). S. I. Cohen and T. N. Jones, A Summary of Density Measurements on Molten Fluoride Mixtures and a Correlation Useful for Predicting Densities of Fluoride Mixtures of Known Compositions, ORNL-1702 (May 14, 1954). W. D. Powers and G. C. Blalock, Heat Capacities of Composition No. 12, No. 44, and of K;CrF ., ORNL CF-54-5-160 (May 20, 1954). S. I. Cohen et al., Pbysical Property Charts for Some Reactor Fuels, Coolants, and Miscellanecus Materials (4th Edition), ORNL CF-54-5-188 (June 21, 1954). S. I, Cohen and T. N. Jones, Measurement of the Density of Ligquid Rubidium, ORNL CF-54-8-10 (Aug. 4, 1954). W. D. Powers and G. C, Blalock, Heat Capacities of Compositions Nos. 39 and 101, ORNL CF-54-8-135 (Aug. 17, 1954). W. D. Powers and S. J. Claiborne, Measurement of the Thermal Conductivity of Molten Fluoride Mixture No. 104, ORNL CF-54-7-145 {(Aug. 24, 1954). W. D. Powers and S. J. Claiborne, Measurement of the Thermal Conductivity of Molten Fluoride Mixture No. 44, ORNL CF-54-10-139 (Ocr. 26, 1954). W. D. Powers and G. C. Blalock, Heat Capacity of Composition No. 40, ORNL CF- 54-10-140 (Oct. 26, 1954). S. I. Cohen and T. N. Jones, Preliminary Measurements of the Viscosity of Composition 20, ORNL CF-55-2-20 (Feb. 2, 1955). S. |. Cohen and T. N. Jones, Measurement of the Viscosities of Composition 35 and Composition 74, ORNL CF-55-2-89 (Feb. 15, 1955). W. D. Powers and G. C. Blalock, Heat Capacity of Lithium Hydride, ORNL CF-55-3-47 (March 7, 1955), S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 72, ORNL CF-55-3-61 (March 8, 1955). S. . Cohen and T. N. Jones, Measurement of the Viscosity of Composition 30, ORNL CF-55-3-62 (March 9, 1955). S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 43, ORNL CF-55-3-137 (March 14, 1955), S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 2, ORNL CF-55-4-32 {April 1, 1955). S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 81 and 82, ORNL CF-55-5-58 (May 16, 1955). S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 78, ORNL CF-55-5-59 (May 16, 1955). W. D. Powers and G. C. Blalock, Heat Capacities of Compositions 30, 31, and 40, ORNL CF-55-5-87 (May 18, 1955). W. D. Powers and G. C. Blaleck, Heat Capacities of Compositions 70 and 103, ORNL CF-55-5-88 {(May 18, 1955). : S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 86, ORNL CF-55-7-33 (July 7, 1955}, H. W. Hoffman, Physical Properties and Heat Transfer Characteristics of an Alkali Nitrate-Nitrite Salt Mixture, ORNL CF-55-7-52 {July 21, 1955). 185 185 W. D. Powers and G. C. Blolock, Heat Capacity of Composition 102, ORNL CF-55-5-8 (Aug. 1, 1955). S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 88, ORNL CF-55-8-21 (Aug. 1, 1955}, S. |. Cohen and T. N. Jones, Measurement of the Viscosities of KBeF, and NaBng and Some Observations on (LiF-BeF,, 50-50 mole %), ORNL CF-55-8-22 (Aug. 1, 1955). S. i, Cohen and T. N. Jones, Measurement of the Viscosity of Composition 70, ORNL CF-55-9-31 (Sept. 6, 1955). S. |. Cohen and T. N. Jones, Measurement of tke Viscosity of Compositions 87, 95, and 104, ORNL CF-55-11-27 (Nov. 4, 1955), S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 90 and 96 and (KE-BeF,; 79-21 mole %}, ORNL CF-55-11-28 (Nov. 8, 1955). W. D. Powers and G. C. Blalock, Heat Capacities of Compositions No. 98 and No. 104, ORNL CF-55-11-68 {Nov. 15, 1955). S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 97 and 98, ORNL CF-55-12-127 {Dec. 23, 1955). S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 31, ORNL CF-55-12-128 (Dec. 23, 1955). W. D. Powers and G. C. Blalock, Enthalpies and Heat Capacities of Sclid and Molten Fluoride Mixtures, ORNL-1956 (Jan. 11, 1956). S. |. Cohen and T. N. Jones, The Effect of Chemical Purity on the Viscosity of a Molten Fluoride Mixture, ORNL CF-56-4-148 (April 17, 1956). S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 12, 14, and 107, ORNL CF-56-5-33 (May G, 1956). W. D. Powers and G. C. Blalock, Heat Capacities of Several Scdium Fluoride, Zirconium Fluoride Compositions, ORNL CF-56-5-66 {May 14, 1956). W. D. Powers and G, C. Blalock, Heat Capacity of Composition No. 100, ORNL CF- 56-5-67 (May 14, 1956). W. D. Powers end G. C. Blalock, Heat Capacity of Composition No. 81, ORNL CF- 56-5-68 (May 14, 1956). RADIATION DAMAGE O. Sisman, LITR Fluoride Fuel Test Loop, ORNL CF-53-9-180 (Sept. 25, 1953). W. E. Brundage and W. W. Parkinson, The Radiation-induced Corrosion of Beryllium Oxide in Sodium at 1500°F, ORNL CF-53-12-42 {Dec. 3, 1953). H. J. Stumpf and B. M. Wiiner, Radiation Damage to Elastomers, Lubricants, Fabrics, and Plastics for Use in Nuclear-Powered Aircraft, ORNL CF-54-4-221 {April 15, 1954). M. T. Robinson, Release of Xenon from Fluoride Fuels: Proposal for an Experimenial Program, CRNL CF-54-6-4 (June 2, 1954). M. T. Robinson, Removal of Xenon from Fluoride Fuels: Preliminary Design of In-Pile Equipment, ORNL CF-54-9-36 (Sept. 3, 1954). M. T. Robinson, Volatilization of Fission Products from Fluoride Fuels, ORNL CF- 541265 (Dec. 10, 1954}, M. T. Robinsen, 5. A. Reynolds, and H. W. Wright, The Fate of Certain Fission Products in the ARE, ORNL CF-55-2-36 (Feb. 7, 1955). M. T. Robinson, A Theoretical Treatment of Xe!33 Poisoning in the ARE and the ART, ORNL CF-55-5-22 (May 2, 1955). . M. T. Robinsen, Deposition of Fission Products from Fluoride Fuels, ORNL CF-55-5-99 {May 16, 1955), M. J. Feldman et al., Metallographic Analysis of Fuel Loop I, ORNL CF-55-4-22 (June 21, 1955). M. T. Robinson, The Fate of Fission Productis in ANP Fluoride Fuels, ORNL CF- 55-7-12 (July 1, 1955), M. T. Robinson and D. F. Weekes, Design Calculations for a Miniature High-Temperature in-Pile Circulating Fuel Loop, ORNL-1808 (Sept. 19, 1955). M. J. Feldman, E. J. Mantheos, and A. E. Richt, Metallographic Analysis of Fuel Loop II — Control Sections from Nose Piece, ORNL CF-55-9-91 (Sept. 26, 1955). M. T. Robinson, A Theoretical Study of Xel33 Poisoning Kinetics in Fluid-Fueled, Gas-Sparged Nuclear Reactors, ORNL-1924 (Feb. 6, 1956). M. T. Robinson and J. F. Krause, The Use of Li in In-Pile Corrosion Testing of ANP Fuels, ORNL CF-56-2-25 (Feb. 7, 1956). | C. Ellis et al., Examination of ANP In-Pile Loop 5, ORNL CF-56-7-48 (JUIY 16, 1956). FUEL RECOVERY C. F. Coleman, Recovery of Uranium as a Single Product from Fluorides, ORNL-1500 (March 31, 1953). A. T. Gresky, E. D, Arnold, and R. J. Klotzbach, Potentialities of Fused Salt—Fluoride Volatility Methods of Processing Irradiated Fuel, ORNL-1803 (March 28, 1956). SHIELDING Report on the ANP Shielding Board, ANP-53 (Oct. 16, 1950). W. .. Wilson and A. L. Walker, An Aircraft Shield Analysis, ORNL CF-52-8-120 (Aug. 20, 1952). E. P. Blizard, Materials Research for Mobile Reactor Shielding, ORNL CF-52-11-129 (Nov. 18, 1952). J. B. Trice, Summary of Results from Recent Fireball Shielding Studies, ORNL CF- 53-6-165 (June 11, 1953). J. E. Faulkner, Critigue of LiH as a Neutron Shield, ORNL CF-53-12-13 (Dec. 2, 1953). A. P. Froas, Shield Designs for the Reflector-Moderated Reactor, ORNL CF-54-2-36 (Feb. 4, 1954). F. C. Maienschein, F. Bly, and T. A. Love, Sources and Attenuation of Gamma Radiation from a Divided Aircraft Shield, ORNL-1714 {Aug. 20, 1954). R. M. Spencer and H. J. Stumpf, The Effect on the RMR Shield Weight of Varying the Neutron and Gamma Dose Components Taken by the Crew and Comparison of the RMR Shield Weight to That for an Idealized Shield, ORNL CF-54-7-1 (Aug. 26, 1954). F. N. Watson et al., Lid Tank Shielding Tests of the Reflector-Moderated Reactor, ORNL-1616 (Oct. 5, 1954). E. P. Blizard, Fraction of Biological Dose Due to Thermal Neutrons in Aircraft Reactor Shields, ORNL CF-54-11-113 (Nov. 19, 1954). J. B. Dee, Jr., and H. C. Woodsum, Shield Weights for the CFRE, ORNL. CF-54-12-154 (Dec. 20, 1954). E. P. Blizard, Spectrometer Measurements of Fission-Product Gamma Rays for the CFR, ORNL CF-55-4-122 (April 21, 1955). J. Van Hoomissen and H. E. Stern, Comparison of Hydrogenous Materials for Use in Aircraft Shields, ORNL CF-55-6-4Q (June 9, 1955}, J. B. Dee, C. A. Goetz, and J. R. Smoclen, Sources of Radiation in a 300-Mev Circu- lating-Fuel Reactor, ORNL CF-55-7-35 (July 12, 1955). J. R. Smolen, Pbysical Properties of Boral Used in Lid Tank Mockups, ORNL CF- 56-1-94 (Jan. 23, 1956). C. D. Zerby, A Monte Carlo Method of Calculating the Response of a Point Detector at an Arbitrary Position Inside a Cylindrical Shield, ORNL-2105 (June 12, 1956). J. R. Smolen, A Preliminary Investigation of B1® Enriched Boron Curtains, ORNL CF- 56-6-163 {June 28, 1956). First Semiannual ANP Shielding Information Meeting, May 7 and 8, 1956, Vols I Through 111, ORNL-2115 (July 2, 1956), 187 188 ORNL-528 ORNL-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 CRNL-1375 ORNL-143¢9 ORNL-1515 ORNL-1556 ORNL-1609 ORNL-1649 ORNL-1692 ORNL-1729 ORNL-1771 ORNL-1816 ORNL-1864 ORNL-1896 ORNL-1647 ORNL-2012 ORNL-2061 ORNL-2106 ORNL-2157 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORTS Period Ending November 30, 1949 Period Ending February 28, 1950 Feriod 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 Periocd 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 Feriod 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