D.N. Sunderman

Diffusion of Xenon in Ceramic Oxides

Postirradiation measurements were made at 700° to 1500°C of the release of 133Xe recoiled into single‐crystal α‐Al2O3, BeO, MgO, and ZrO2 crystals during neutron irradiation in contact with UO2 powder. The release pattern from most of the specimens was identical: a high rate of loss of 133Xe for several minutes followed by a slower release for which the cumulative fractional release was linear with the square root of the heating time. Diffusion coefficients for 133Xe have been calculated from the slope of the release curve following the initial release. The magnitudes of the diffusion coefficients above 1100°C for α‐Al2O3, BeO, and MgO were nearly the same whereas the diffusion coefficients in ZrO2 were ten to a hundred times larger. The activation energy for 133Xe release from α‐Al2O3 above 1100°C was 64 kcal/g‐mole. The xenon release behavior has been compared to anion diffusion in these oxides.

Fission fragment induced expansion in ceramic materials

Abstract The fission fragment induced expansion of ceramic materials was studied by recoiling fission fragments into one face of thin, flat ceramic specimens and then using interferometry to measure the specimen deflection. Irradiation caused the specimens to become spherically curved and the fission-fragment-induced expansion increased in the order : MgO SiO 2 BeO Graphite Al 2 O 3 ZrO 2 Pyrolytic Carbon SiC No difference in expansion was observed for singlecrystal and polycrystalline specimens. Expansion normal to the surface was measured for single crystal α-Al2O3 by recoiling fission fragments into a portion of the surface and then measuring the expansion step at the boundary between irradiated and unirradiated regions of the surface. The expansion was found to be independent of crystallographic orientation. Radiationinduced expansion normal to the surface was at least three times the expansion in the specimen plane, indicating that fission fragment induced stresses were preferentially relieved by expansion normal to a free surface. Damage annealing during postirradiation heating was measured and approximate values for radiation expansion coefficients and stresses in irradiated specimens were calculated.

AN IMPROVED TECHNIQUE FOR THE DETERMINATION OF SHORT-LIVED FISSION-PRODUCT GASES

Radioactive gases are passed longitudinally through a long chamber containing an electrically charged stainless steel rod. When one of the gas nuclei decays in the chamber, the daughter nucleus is drawn to the rod. After a given period of exposure to the gas, the rod is sectioned and examined by gamma spectrometry. The short-lived nuclides Kr/sup 89/, Kr/sup 91/, Kr/sup 92/, Xe/ sup 138/, Xe/sup 139/, Xe/sup 140/, and Xe/sup 141/ are determined in the gas by this method, by means of determinations of the activities on the rod of their daughters Rb/sup 89/, Sr/sup 91/, Sr/sup 92/, Cs/sup 137/, Cs/sup 138/, Ba/sup 139/ Ba/sup 140/, and La/sup 141/ respectively. The distributions of the decay activities along the rod agree with theoretical predictions. (T.F.H.)

DEVELOPMENT OF A FUEL-ELEMENT LEAK-DETECTION SYSTEM BASED ON THE PRINCIPLE OF ISOTOPIC EXCHANGE

The selective removal of halide fission products from an aqueous solution by exchange with the halide in a solid silver halide was studied as the basis for a fuel-element leak detector. The retention of fission-product halides on a silver halide column was investigated as a function of coolant flow rate, halide anion, and column size. Fission prcduct decontamination factors and predicted operating lifetimes were obtained for a number of reactor operating conditions. It is concluded that a sensitive, rapid leak detector for a water- cooled reactor could be constructed from a silver bromide or iodide column monitored by a neutron detector to detect delayed neutrons from the halide fission products. The feasibility of gross gamma monitoring was found to be dependent upon the intensity of the gamma background arising from absorbed fission products on the silver halide column. (auth)

RADIOMETRIC METHODS FOR THE DETERMINATION OF MAGNESIUM AND CALCIUM IN PORTLAND CEMENT

Radiometric methods of analysis for magnesium and calcium have been developed as part of a program for the U. S. Atomic Energy Commission. Office of Isotopes Development, which are applicable to the determination of these elements in portland cement Both methods employ, as a precipitant, a standard solution of (NH/sub 4/)/sub 2/HPO/sub 4/ labeled with phosphorus-32. In the presence of NH/ sub 4/OH, this reagent precipitate; MgNH/sub 4/PO/sub 4/ or Ca/sub 3/(PO/sub 4/)/ sub 2/ from a solution of magnesium or calcium ions. The reduction in the radioactivity level of the labeled phosphate solution after precipitation serves as a measure of the phosphate reacted and thus a measure of the quantity of magnesium or calcium present. Studies have been made of the effects of reagent concentration, NH/sub 4/OH concentration, and other experimental variables. The interference of other elements present normally in portland cement and its raw materials has been determined. The concentration ranges for highest accuracy have been found to be 5 to 15 mg of MgO per 100 ml and 15 to 30 mg of CaO per 50 ml. (auth)

Apparatus for the Study of Fission-Gas Release From Fuels During Postirradiation Heating at Temperatures Up to 1600 C

An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)

MEASURING THE RELEASE OF SHORT-LIVED FISSION GASES DURING CAPSULE IRRADIATIONS

A technique is described for the determination of the release of short- lived fission gases during capsule fuel irradiations. Sweep helium passes over the fuel specimen and carries fission gases to a delay trap located close to the capsule. Short-lived fission gases decay in the trap and deposit long-lived daughters which are assayed by radiochemical methods after trap removal. In two demonstrations of the technique the radiochemical analyses have provided information on the rate of release of 1.7-sec Xe/sup 141/, 16-sec Xe/sup 140/, 3.9 min Xe/sup 137/, and 3.2-min Kr/sup 89/. The technique should find considerable application in the irradiation evaluation of fuel materials when it is necessary to determine the contribution shont-lived fission gases make to the coolant-contamination problem. (auth)

APPARATUS FOR THE STUDY OF FISSION-GAS RELEASE FROM NEUTRON-ACTIVATED FUELED GRAPHITE

A simple laboratory apparatus for the study of fissiongas release from neutron-activated fueled graphite was developed. Xenon-133 released from a heated specimen is carried in a helium sweep gas to a charcoal trap, where the accumulated activity is monitored continuously by a scintillation detector, ratemeter, and pen recorder. The maximum specimen temperature (2500 deg F) is achieved in 10 min with an induction heater. All instrumentation is commercially available. Data for several neutron-activated fueled-graphite specimens heated in the range from 800 to 2500 deg F are presented to illustrate the typical results obtained with the apparatus. (auth)