Production from •'-'^B in control rods and poisoned spines 6.1.5 Summary of tritium sources 6.2 Estimated Tritium Inventories in Core 2 Components Based on Measured Concentrations 6.2.1 Fuel inventory 6.2.2 Sleeve and spine inventories 6.2.3 Removable radial reflector inventory 6.2.4 Inventories in the upper and lower axial reflector regions of the driver fuel elements 6.2.5 Summary of Inventories derived from observed specific activities 6.3 Estimated Tritium Losses from the Primary System 6.3.1 Transport to the fission product trapping system 7 6.3.2 Transport to the chemical cleanup system (and disposal as solid waste) 6.3.3 Leakage through the steam-generator tubing 6.3.4 Loss due to containment leakage 6.3.5 Discharge with gaseous waste flow 6.3.6 Discharge with liquid wastes 6.

References for Section 1 8 2. GAMMA SCANS OF PRIMARY LOOP PIPING 8 2.1 Primary Circuit Description and Scanning Locations 8 2.2 Detection Equipment and Calibration Methods 11 2.3 Data Summary of Primary Loop Gamma Scans l6 2.k Discussion of Primary Loop Gamma Scan Data 23 References for Section 2 26 3. PRIMARY COOLANT SAMPLERS FOR GASEOUS RADIONUCLIDES AND PARTICULATES 26 3.1 Description of Experiment 26 3.2 Tabulated Results ^0 3.2.1 Radionuclide inventories on sample components .... ^-0 3.2.2 Dust collected on impactor stages and filters .... ^-0 3-3 Axial Distributions in Sampler Diffusion Tubes and Nozzles 65 3 .3 • 1 Diffusion tubes 65 3.3*2 Axial distributions on sampler nozzles 67 3.h Discussion of Diffusion Tube Results 77 3A.I Theoretical behavior of diffusion tubes 77 3.U.2 Effect of the nozzle on coolant sampler performance 79 3.^-.3 Axial distribution of nuclides on the sampler diffusion tubes 8l References for Section 3 88 k. FISSION GAS CONCENTRATIONS IN THE PRIMARY COOLANT AND CORE PURGE GAS 89 k.l General Description of Fission Gas Experiment 89 k. 2 Experimental Methods 91 k. 2.1 December 1970 experiment 91 ^-.2.2 February 1973 measurements 9h k.2.3 September 1973 measurements 98 k.3 Release-to-Birth Rate Ratios of Fission Gases 108 k.h Discussion of Fission Gas Determination 109 References for Section k 113 5-ANALYSES OF DUST COLLECTED IN THE CYCLONE SEPARATORS AND FROM THE EXTERIOR SURFACE OF THE FUEL ELEMENTS 113 5.1 Dust Collection Method 113 5.2 Summary of Cyclone Separator Dust Data 117 5.3 Density Separation of Sample D3 by Zonal Centrifugation . 12^ 5.3.1 Top fraction 128 5.3.2 Second fraction 129 5.3.3 Third fraction 129 5.3 .k Fourth fraction 129 5.h Scanning Electron Microscope (SEM) Photographs of Particles in Samples D3 and D4 Size Fractions 130 5.5 Further Particle Size Distribution Data Emphasizing 0.1 to 10 urn Size Range ll+8 5.6 Characterization of "Soot" Collected from Exterior Surface of Fuel Elements 153 5.7 Dust Concentration and Production Rate in the Primary System 159 References for Section 5 l6l APPENDIX A: DOCUMENTATION OF GAMMA SPECTROSCOPY METHODS 163 Al. Reference Nuclear Data Source 163 A2. Measurement of Photopeak Count Rates and Energies 163 A2.1 Peak finding and single peak measurement 163 A2.2 Analysis of multiplet peak regions l66 A3. Counting Efficiencies 168 A3.1 Usage of counting efficiencies l68 A3-2 Experimental determination of counting efficiencies 169 A3.2.1 Hot-valve cooling line 170 A3 .2.2 Primary circuit piping 170 A3.2.3 Gaseous diffusion tubes 172 A3.2 .k Purge gas vials and flow cells 172 kk. Formula Used to Compute Activities and Fission Gas Concentrations 173 References for Appendix A 1

DISCLAIMERGindler'') lists eleven methods and reports on the range of quantity of uranium in micrograms for which each method is applicable. These methods and the lower limits of analysis are listed in Table I. An alternate approach to the neutron activat'ion analysis of uranium is to count the delayed neutrons emitted by the fission products of U-235. Echo and Turk(6) first described the application of this technique to the determination of By a systematic irradiation and counting procedure, they made determinations of u~~~ in a number of liquid samples containing known q~antities of which -2 varied from 10 to. lom3 micrograms. Three synthetic ore samples containing known amounts of $35 were also analyzed. They reported 0.05 microgram of u~~~ as a lower limit for a determination. They found the method to be rapid, and accurate, to require very little sample preparation, and to be free of any interferences except from those materials which have high neutron fission cross-sections (such as ?u 239 and $33) or high-neutron capture cross-sections (such as mie el'^) has made a more extensive study of delayed neutron emission for the determination of fissionable materials. He applied the method.to the determination of uranium in a large number of geolog'ical'~samp1es and to samples of pure uranium reagents, as well as the determination of the isotopic composition of a uranium sample. from the fission of $35, $33, and Pu239 by both thermal and fast neutrons and from the fission of U238 and d32 by fast neutrons.It was found that the delayed-neutron activity resulting from the fission of each of these nuclides could be resolved into six groups with fair agreement in the half-lives of corresponding groups.Slight differences in the half-lives of corresponding groups were obtained when-different nuclides were fissioned as well as when the method of fission was different, as for example, when U-235 was fissioned first with thermal neutrons and then with fast neutrons.Quite.large differences between corresponding groups for absolute group yields were obtained when different nuclides were fissioned.The data obtained, by Keepin, et a1, (9) for thermal neutron fission of 335, $33, and l?u239 is given in Table 111 Table IV lists An analysis for uranium is made by irradiating a test sample in a nuclear reactor for a period of time ranging from seconds to minutes, rapidly removing the sample to a neutron counting facility, allowilig b~b s o l u t e group y i e l d (dp) means t h e number of delayed neutrons per 100 f i s sians .t h e counter t o c o l l e c t counts f o r a period of time and recording t h e ' c o l l e c t e d count, A blank sample and a comparator sample a r e then t r e a t e d p r e c i s e l y a s was t h e t e s t sample. The blank sample u s u a l l y c o n s i s t s of an empty r a b b i t . In those cases i n which a uranium a n a l y s i s i s desired on a material t o which uranium has been added, t h e o r i g i n a l matrix material t o which no uranium has been added can serve a s the blank along wi...

No abstract

This report describes the postlrradiation examinations of driver fuel element EO6-OI, which had been irradiated an equivalent of 38^ full-power days in Peach Bottom, Unit 1. The fuel element is described in detail and its temperature and irradiation service history briefly outlined. Results presented include: (l) visual observations; (2) critical dimensions of fuel compacts, sleeve, and spine: (3) axial distributions of gamma-emitting nuclides plus H and Sr; {k) radial distributions of these nuclides in the sleeve and spine at three axial locations in the fueled regions and three locations in the upper reflector; (5) metallographic examination of samples of fuel compact material; and (6) burnup determinations via radiochemical analyses at two compact locations.

This document contains information of a preliminary nature. It is subject to revision or correction and therefore does not represent a final report.

No abstract

Uranium and thorium were measured by absolute neutron activation analysis in high-pcrity materials used to manufacture semiconductor memories. The main thrust of the study concerned aluminum and aluminum alloys used as sources for thin film preparation, evaported metal films, and samples from the Czochralski silicon crystal process. Average levels of U and Th were found for the source alloys to be ^65 and ^45 ppb, respectively. Levels of U and Th in silicon samples fell in the range of a few parts per trillion. Evaporated metal films contained about 1 ppb U and Th, but there is some question about these results due to the possibility of contamination. *0perated by Union Carbide Corporation for the U.S. Department of Energy, under Contract No. W-7405-eng-26.