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ANP PROJECT PROGRESS REPORT SECTION 6-e 1 0 1 2 3 4 S 6 7 P !XALE IN INCHES T SUPPORT ASSEMBLY- REMOVABLE PLUG- SECTION A-A ( 0 1 2 3 4 5 6 7 P SCALE IN INCHES Ee ISLAND FUEL Be rcerrrr 2-01-059-79 EORAL INCONEL (SMC PRESSURE SHELL) SALT REGION INCONEL EORAL INCONEL (CFRMR PRESSUfi I€ SHELL) INCONEL ALKYL BENZENE CONTAINER K- -2 SCALE IN FEET Fig. 5.4.1. The Shield Mockup Core. i .
an lnconel shell and a complicated boron-copper layer. In considering the radiation as seen outside the shield, this region can be adequately simulated in the SMC with a shell of Intonel followed by layers of bora1 to give the proper density in g/cm2, Calculations are being carried out at Pratt & Whitney Aircraft to determine the importance of each region as a source of radiation. The heat exchanger region just beyond the first boron curtain affects the crew compartment dose rate in several ways: (1) it attenuates the radiation from the core and the beryllium; (2) it is a source of some capture gamma radiation from core neutrons; and (3) it is a source of delayed neutrons and fission-productdecay gamma rays. In order to account for the first two effects, the heat exchanger region will be mocked up with fused salts (in the form of NaF and KF) and NaK, as a homogenized mixture. This will be accomplished by heating the salt, the NaK, and the can to about 1000°C in an evacuated furnace. The cans will be placed in two layers to eliminate leakage paths between the cans. The lnconel in the heat exchanger will be split to form two shells around the salt region. The inner shell will act as the pressure shell for the SMC. Outside the outer heat exchanger shell will be the second sodium-cooled boron curtain, which, like the first curtain, will be mocked up with boral. The regular CFRMR pressure shell will follow the heat exchanger region. It has bee0 proposed that part of this shell be split off for use in mount- ing the lead shielding. This section is to be removable to allow the lead shield to be changed without dismantling the reactor, The neutron shielding material will be contained in an aluminum tank. The optimized neutron shield will be a sphere placed off-center with respect to the reactor. It is proposed to permit lateral motion of the neutron shield, while using water as shield material, to check the present optimization. Later the shield container is to be sealed in the optimized position, and neutron shielding materials other than water can be used. In order to facilitate the measurements at thhtTSF the whole reactor and shield system has been de- signed so that it can be rotated about the vertical axis. COMPARISON OF SMC AND CFRMR An examination of some of the results of the calculations performed by Pratt & Whitney indicates PERIOD ENDING JUNE 10, 1956 how closely the SMC radiation simulates the CFRMR. Thermal-neutron captures in the reflector and the power distribution within the core have been considered, and the importance of each region of the reactor as a gamma-ray source is being investigated. Neutron Captures in Beryllium Since approximately 20% of the dose rate in the crew compartment is expected to originate from neutron captures in the beryllium, this source is to be accurately simulated. Figure 5.4.2 shows the captures that can be expected in the SMC beryllium with normal water as the coolant in the core. The lower curve shows the absorptions in the SMC beryllium when the space between the fuel plates is completely filled with normal water and the reactor is operated at room temperature. oE8Rc+ 2-09-059-776 CONFIGURATION 160: SPACE BETWEEN FUEL PLATES FILLED WITH 100 % H20 CONFIGURATION 167: SPACE BETWEEN FUEL PLATES FILLED WITH 50% H20-50% AI CONFIGURATION 168: SPACE BETWEEN FUEL PLATES FILLED WITH 25%H20-75%AI CONFIGURATION 168A: SPACE BETWEEN FUEL PLATES FILLED WITH 12.5% H20--87.5% AI 35 40 45 50 55 60 65 SPACE POINTS IN REFLECTOR Fig. 54.2. Comparison of Neutron Captures in Beryllium Reflector of SMC for Various Configu- rations with Neutron Captures in Reflector of CFRMR. 281
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Part 1 AIRCRAFT REACTOR ENGINEERING
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STATUS OF ART DESIGN Design work on
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of pressure loads. The idealized mo
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UPPER DECK 7 @ PRESSURE SHEL PERIOD
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SOD,UM r--- INCONEL TANK ROOF 0 0.5
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TABLE 1.1.1. NaK PUMP SPEEDS AND HO
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62 CALCULATED HEATING (w/crn3) N w
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where I = radius of source (taken a
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heat exchanger was used. The heatin
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head which are bounded by various t
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h bd * W - EcBRc+ ORNL-LR-DWG I4918
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PERIOD ENDlNC JUNE 10. 1956 TABLE 1
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ENRICHER ACTUATOR J. M. Eastman Spe
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hd - Thus, even with an input signa
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- . 140 120 p 100 g d 80 8 L s In y
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and lose their hardness in the pres
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p. W 2.5 0.5 0 400 OPERATING TIME (
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i 20 too 80 40 20 PERIOD ENDING JUN
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c' I - + r 500 450 400 3 50 300 2 2
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01) 0V3H 5: N 0 2 s 0 0 0 5: 0 2 s
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the first 200OF and B0F/sec for the
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PERIOD ENDING JUNE 10, 1956 TABLE 1
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L 0 _, I, -* t; - PERIOD ENDING JUN
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LJ 0.005 in. on the diameter and a
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I 3 -I W CALROD HEATER 1500 I 1400
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-17 Inconel I .._I_ I" .. _ _ ~ ._
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i E ART FACILITY F. R. McQuilkin Co
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L7i PERIOD ENDING JUNE 10, 1956 Fig
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PERIOD ENDING JUNE 10, 1956 Fig. 1.
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ART OPERATING MANUAL W. B. Cottrell
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Part 2 CHEMISTRY W. R. Grimes
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W 2.1. PHASE EQUILIBRIUM STUDIES' C
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- 0 Q 3 G 4000 900 800 700 a w a 5
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- a t a w 1000 900 000 - P 700 3 2
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- 0 e 4 000 900 800 5 700 I- a a W
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ZrF4 9t2 PERIOD ENDING JUNE 10, 795
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U c * NaF*BeF2-2NaF*BeF2 triangle c
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THE SYSTEM MgF2-CaF2 L. M. Bratcher
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2.2. CHEMICAL REACTIONS IN MOLTEN S
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I id PERIOD ENDING JUNE 10, 1956 an
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PERIOD ENDlNG JUNE 10, 1956 V which
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u the salt-metal interface, and the
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of this reaction will be made at th
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+- + z 6 e 6 5 3 3 > k i m 3 2 2 1
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FeF,. The FeF, saturation points we
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UNIF, = yN should be unity (a pure
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Although it appears that the solubi
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, . 2.3. PHYSICAL PROPERTIES OF MOL
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-0 a rn u) u) DO2 oot c ;o rn OS g
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IL, - PERIOD ENDING JUNE 70, 7956 O
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8 00 700 600 - 0 0, w 500 a 3 I- U
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supporting plaque is loaded into th
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u 2.4. PRODUCTION OF FUELS c G. J.
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half of fiscal year 1957. This esti
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2.5. COMPATIBILITY OF MATERIALS AT
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volume of 1 M tartaric acid solutio
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After the trap has been opened, bot
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\ Part 3 METALLURGY W. D. Manly
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FLUORIDE FUEL MIXTURES IN INCONEL F
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PERIOD ENDING JUNE 10, 1956 TABLE 3
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Fig.3.1.3. Region of Maximum Attack
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(d . terminated after 1217 and 1339
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- . LJ BRAZING ALLOYS IN LIQUID MET
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PERIOD ENDING JUNE 10, 1956 NaK wer
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I PERIOD ENDING JUNE 10, 1956 I Fig
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0. PERIOD ENDING JUNE 10, 1956 Fig.
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Fig. 3.2.10. Apparatus for Studying
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STATIC TESTS OF INCONEL CASTINGS R.
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t (JnOOSll 3n918-3NOt IOH (JoOOSI)
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after the l00hr test. The extent of
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tests of the Lindsay Mix specimens,
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LJ DEVELOPMENT OF NICKEL-MOLYBDENUM
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LJ PERIOD ENDING JUNE 10, 1956 Fig.
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2. Canning of the billets with '/,-
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165 . c, e . c3 a PERIOD ENDING JUN
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u allow the billet to start through
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NEUTRON SHIELD MATERIAL FOR HIGH-TE
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PERIOD ENDfNG JUNE 10, 7956 Fig. 3.
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wrought plate used as base material
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J point. A mixture of 50-50 vol % L
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WELDING PROCEDURE: VERTICAL FIXED,
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4 2 t 4 2 in. t 2 WCLASSFKO ORNL-LR
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FABRICATION OF JOINTS BETWEEN PUMP
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* h WELDING PROCEDURE PERlOD ENDING
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1 1 6 in. PUMP BARREL t SECTION AA
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'4 x 6 x 20-in. INCONEL PLATES (/*-
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UNCLASSIFIED PHOTO 17248 Fig. 3.4.1
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through the radiator by passing col
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L, . 1 Q t V ERIOD ENDING JUNE 10,
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to permit the examination of indivi
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LJ . t L t * I LJ Fig. 3.4.33. Tube
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Fig. 3.4.37. Inner Surface of Tube
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EFFECTOF ENVIRONMENTONCREEP- RUPTUR
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PERIOD ENDING JUNE 10, 1956 UNCLASS
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Fig. 35.7. Surface Effect on the St
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44,000 42,000 40,000 - ._ (D 8000 v
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\:r 1 U UNCL 1158 W ioo,ooo 80,000
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fuel mixture (No. 30) NaF-ZrF -UF,
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i u * ‘*$ , .\. The Lindsay Chemi
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6, * c L . c e 4 gi depths greater
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i 4 Part 4 HEAT TRANSFER AND PHYSIC
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4.1. HEAT TRANSFER AND PHYSICAL PRO
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This relationship gives the ratio o
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The parameters of the’experiment
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t Fig. 4.1.7. Fuel Annulus of the V
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uncooled wall temperature that woul
Page 215 and 216: SHIELD MOCKUP REACTOR STUDY L. C. P
Page 217 and 218: Liquid (688 to 8986C) I . HT - H3Q
Page 219 and 220: THERMAL CONDUCTIVITY W. D. Powers T
Page 221 and 222: cd Fig.4.2.4. Slingers from MTR In-
Page 223 and 224: a plug when they came in contact wi
Page 225 and 226: couples were used on this loop to e
Page 227 and 228: 0 20 40 60 80 lo0 I20 440 (60 180 2
Page 229 and 230: "fast" neutron in a graphite reacto
Page 231 and 232: eached thermal equilibrium in an in
Page 233 and 234: 0.5 0.4 - a3 l- W z 0.2 0.1 UNCLASS
Page 235 and 236: TABLE 4.2.3. RESULTS OF EXAMINATION
Page 237 and 238: U could be evacuated was used for t
Page 239 and 240: ! j * . x - iu I PERlOD ENDlNG JUNE
Page 241 and 242: CRITICAL EXPERIMENTS FOR THE COMPAC
Page 243 and 244: . I) a W PERIOD ENDING JUNE 10, 195
Page 245 and 246: U I I R i c . --. in. long, 18’4,
Page 247: I, . R t 3 4 . C 7 6 5 W 5 a c3 E4
Page 250 and 251: 1 a
Page 252 and 253: ANP PROJECT PROGRESS REPORT ot(E) =
Page 254 and 255: ANP PROJECT PROGRESS REPORT TABLE 5
Page 256 and 257: ANP PROJECT PROGRESS REPORT TABLE 5
Page 258 and 259: 6 ANP PROJECT PROGRESS REPORT 2 lo7
Page 260 and 261: ANP PROJECT PROGRESS REPORT GAMMA-R
Page 262 and 263: ANP PROJECT PROGRESS REPORT 5.3.3,
Page 264 and 265: ANP PROJECT PROGRESS REPORT A I P I
Page 268 and 269: ANP PROJECT PROGRESS REPORT .. .._
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