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SOURCE Fig. 6.8. Geometry for Calculating With the obvious substitutions, the dose at a distance 2 with a void of thickness Z - 2’ becomes 1 f 2d Dv(2) = D(Z’ 1 1t-2a’ where Dv(Z) is the dose with a void and D(Z) is the dose with water alone. The curve in Fig. 6.7, labeled ‘ho transition section:’ was calculated by using this formula. The experimental points are also indicated, and it can be seen that the agreement between experimental data and the calculations is excellent. Some very rough calculations have been made to explain the effect of the transition sections on the center-line data. If the water sections are assumed to be opaque to neutrons, the calculated dose is lower than the observed dose. More accurate calculations are yet to be made. PERIOD ENDING DECEMBER 10, 1952 ViRTUAt SOURCE wmw DWG. 17371 DETECTOR the Effect of a Void on Neutron Uose. G-E INLET AIR DUCT A mockup of the G-E inlet air duct has been tested in the Lid Tank Facility. Some gamma shielding in the form of slabs of 7/8 in. of iron and 1 1/2 in. of lead was used to simiilate the design shield along the center line. In spite of care in choosing from available slabs of iron and lead, it was impossible to make the mockup simulate the actual situation very closely. As a consequence, caution should be used in applying the gamma data to the design. Thermal- and fast-neutron and gamma data were taken with the standard Lid Tank instruments, and the results have been converted into isodose plots, Figs. 6.9 and 6.10. It is unfortunate that the source strength was completely inadequate to enable measurement of the fast-neutron dose at the end of the duct. The final step in this experiment was to study the attenuation of the 71
- *0° 1 BO 440 120 , '0° i 80 , I 40 'y L 20 L 0 I60 140 120 100 80 60 40 20 -3 v. HORIZONTAL DISTANCE FROM SOURCE AXIS(cm1 THERMAL NEUTRON FAST NEUTRON DWQ. I71 59 1 :N mrep, 'hr -20 -40 -60 -80 -100 -120 -140 -160 Fig. 6.9. Neutron-Isodose Measurements Around G-E Inlet Air 5uct lockup. TO
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Reports previously issued in this s
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e e e PAGE Moore Nullmatic pressure
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.. CrF. .Na. CrFd .Li. CrF6 .......
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PAGE Melting point of 60% Pd-40% Ni
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ANP PROJECT QUARTERLY PROGRESS REPO
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AYP PROJECT QUARTERLY PROGRESS REPO
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in the hot condition corresponds to
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c-' w FILL TANKS LINE VOLUME APPROX
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the disintegration rates and energi
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2. EXPERIMENTAL REACTOR ENGINEERING
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design of the first combination sea
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Page 72: T BLE 5. RADIAL PmITxm- (in. 1 REAC
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ANP PROJECT QIJARTERX,Y PRQCRESS RE
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ANP PROJECT QUARTEHLY PROGRESS REPO
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ANF PROJECT QUARTERLY PROGRESS REPO
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ANP PRQ JECT QUARTERLY PROGRESS RQE
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134 = e . , n N N N - Y) o m -J .m
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constituent was found in this loop.
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(mil SId: 3 6 10 15 20 25 30 Extern
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ANP PROJECT QUARTERLY PRQGIXSS HEPO
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ANP PROJECT QUARTERLY PRQGRESS REPO
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ANP lJHO JECT QUARTERLY PHOGRESS RE
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AhP PROJECT QUARTERLY PHOGRESS REPO
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ANP PROJECT QUARTElilLY PROGRESS RE
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BRAZING ALLOY Nic robr az 60% Mn-40
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ANP PRQJECT QUARTERLY PROGRESS REPO
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ANP PROJECT QUARTERLY PIJQGRESS REP
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ANP PRQJECT QUAMTERLY PWOd;l"aESS R
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on air radiations and simulated fla
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184 Fie. 13.6. Pyrex Thermal Convec
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ANP PROJECT QU.4RTEtlLY I'ROGRESS R
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SUMMARY AND INTRODUCTION The list o
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CF-52-11-85 ORNL- 1 43 3 OWL- 137 2
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H. Seal and Pump Development PERIOD
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12. 13. 14. 15. 16. 1.7. 18. 19. Pr
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i 3. U0,-NaOH Slurry Loop 4. Operat
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7. Fuel Spherical Particle Producti
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