Robert V. Strain

Residual stress measurements in irradiated solution-annealed type 304 stainless steel tubing

Abstract Immersion density and residual stress measurements were made on solution-annealed type 304 stainless steel capsule tubing irradiated up to fluence levels of 9 × 10 22 n/cm 2 ( E > 0.1 MeV). The measured residual stress is dependent on the competition between differential swelling which builds up differential stresses, and irradiation creep which relaxes these stresses. The measurements were analyzed using a bilinear swelling equation formulated with swelling data from the same heat of material. The temperatures and fluence levels of the swelling and slit tube data were each calculated with the same computer code. At high fluence, when swelling was in the steady-state region, the effective irradiation creep rate increased by a factor of about three. Further analysis was made assuming that stress-enhanced swelling and swelling-enhanced irradiation creep were the enhanced relaxation mechanisms.

Fuel Relocation Mechanisms Based on Microstructures of Debris

Argonne National Laboratory (ANL) has performed a number of examinations to determine the microstructure and micro-chemistry of samples of debris from the TMI-2 reactor. These examinations have been a small part of the overall effort to gain an understanding of the TMI-2 accident. As a result of these overall efforts, a general scenario of the response of the core components has been established. In this paper we will describe the microstructure and micro-chemistry of debris from the lower plenum of the reactor and relate these data to a segment of the general scenario dealing with the relocation of this material. The primary tools used at ANL for the examination of material from the TMI-2 core were optical microscopy, scanning electron microscopy and Energy Dispersive X-Ray Spectroscopy, and Scanning Auger Spectroscopy. In some cases these techniques were augmented by the use of gamma spectroscopy, autoradiography, and X-ray diffraction analysis.

Behavior of breached mixed-oxide fuel pins during off-normal high-temperature irradiation

This paper reports on a test containing 19 mixed-oxide fuel pins that was operated in the Experimental Breeder Reactor II (EBR-II) at peak cladding temperatures near 800{degrees} C. Two test pins that had been designed to fail at {approximately}5 at.% burnup and two low-burnup environmental pins failed and then were operated in the run beyond cladding breach mode for 22 days. Very high delayed neutron signals occurred during the irradiation of the test, and it was terminated as a result of high delayed neutron signals and evidence of plutonium in the coolant. Each of the four pins exhibited multiple breaches in the upper half of the fuel column. Measurements of fuel trapped on the filter section of a deposition sampler that was located above the test indicated that {approximately}2.7 g of fuel was lost during the irradiation. Postirradiation examination of the pins indicates that most of the fuel was lost from a single pin. The fuel loss resulted in an increase in the background delayed neutron signal but had no other deleterious long-term effect on the operation of the EBR-II.

Sinusoidal source perturbation experiments with a breached fuel subassembly in the experimental breeder reactor II

A bifrequency reactivity oscillation procedure (ROP) was devised at the Experimental Breeder Reactor II (EBR-II) to be used as a diagnostic tool for characterizing mechanisms responsible for the release and transport of short-lived fission products from the surface of exposed fuel. A series of ROP experiments was conducted during operation of 74% of full power with a breached fuel pin in the core. Detailed analyses of the results using bivariate spectral decomposition and cross-correlation techniques are presented. Comparison of the results of these experiments with those obtained from earlier tests with an unclad fuel source provides conclusive evidence that all nonrecoil fission product release phenomena originate from mechanisms acting inside the breached element itself. In this paper implications of the findings from this study in terms of the goals of high-sensitivity fission product surveillance are discussed.

Fast flux irradiation tests performed at high temperature

Abstract The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O 2 and have depleted UO 2 lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F). The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described. The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment. Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10 22 n/cm 2 show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule. The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.

Transient behavior of “high-swelling” EBR-11 Mark-1A driver fuel in treat

Abstract Transient studies on “high-swelling” Mark-1A driver fuel indicated that temperatures at clad failure, even with irradiated elements, were not significantly lower than clad failure temperatures for normal low-swelling fuel. These temperatures represent thresholds far in excess of the temperatures conceivable for postulated overpower or loss of coolant conditions in EBR-11. In addition to high-speed color photography and cladding temperature measurements, the test capsule was outfitted with electromagnetic motion transducers to follow pre-failure fuel movements. This instrumentation proved to be a powerful tool, both in the determination of the prompt negative feedback reactivity available from fuel expansion, and the description of clad failure phenomena. The nature of cladding failure was similar to observations on normal fuel. Two possible mechanisms for the release of stress from the restricted expansion of sodium trapped between the fuel andcladding appear to be either (1) motion of fuel axially to break the weak interaction forces between the fuel and cladding, or (2) compression of the highly swollen fuel.

The microstructural and microchemical characterization of samples from the TMI-2 core

Abstract Samples of materials from various regions of the TMI-2 reactor core and vessel have been examined at Argonne National Laboratory with a variety of microanalytical techniques. The purpose of these examinations is to characterize the microstructure and microchemistry of the materials so that their origin could be determined, their fission-product content evaluated, and their role in the accident scenario assessed. Macroscopic and microscopic composition inhomogeneities in melted fuel from different reactor locations indicate different cooling rates and solidification temperatures. The mobility of molten fuel could have been enhanced by a low temperature eutectic in the FeCrO system. Stainless steel-clad AgInCd control rods could have failed from a eutectic reaction between the Zircaloy guide tubes and the cladding. Significant concentrations of fission products were not found, but their release from the fuel did not appear to be enhanced by gas-generated channels along grain boundaries.