Albert J. Machiels

Concentration levels of solutes in porous deposits with chimneys under wick boiling conditions

Abstract Concentration levels of highly-soluble impurities and additives in porous deposits with chimneys, when wick boiling is the major mode of heat transfer, have been investigated by a two-dimensional model. Wick boiling promotes the development of high concentration levels of solutes within porous deposits, especially in the region near the intersection of the heating surface and the chimney wall. The maximum concentration factor increases with decreasing porosity; increasing chimney population density; increasing system pressure in the range of interest to LWRs; and approximately exponentially with heat flux and crud thickness. The two-dimensional solute concentration distribution in porous deposits is consistent with limited experimental observations of higher concentration at the chimney wall.

Hydrogen water chemistry for BWRs

Abstract Intergranular stress corrosion cracking (IGSCC) has been responsible for more than 1,000 cases of cracking in austenitic stainless steel piping systems in boiling water reactors (BWRs). This paper presents the status of efforts in the United States to prevent IGSCC in BWRs during power operation by modifying the chemistry of the reactor water. The technical basis for this alternative water chemistry, called hydrogen water chemistry (HWC), is described and the results are presented of an ongoing in-plant program, at Commonwealth Edisons Dresden-2 plant, to verify the HWC concept and systematically assess the consequences of using it in an operating BWR. In addition, progress toward implementation of HWC at other U.S. plants is summarized.

Characterization of UO2 irradiated in the BR-3 reactor

Abstract In order to better understand the behavior of irradiated UO 2 , six fuel rods containing fuel pellets fabricated by wet and dry processes were irradiated under various power histories in the BR-3 PWR (Mol, Belgium). Subsequent to the irradiation, extensive hotcell measurements were performed on complete rods, whole pellets, pellet cross sections or thin slices, and pellet fragments. The results of these measurements collectively underscore several key characteristics of high burnup (37–65 GWd/t) fuel. Although a clear separation of burnup effects and fuel temperature history effects was not possible, the pellet microstructure evolution can be defined in terms of four distinct radial zones. Further, fission gas release increased with burnup but grain growth was not found to be an integral part of the release mechanism. The onset of accelerated Xe release, as indicated by radial profiles by EPMA, corresponded well to a sudden increase in intragranular pore density within a narrow radial zone. Acicular precipitates having apparent similarity to U 4 O 9 phase were found in etched as well as fractured surfaces in high burnup specimens. However, the phase structure was not confirmed and the precipitates might have originated due to changes in the fuel chemistry rather than stoichiometry. Finally, extensive fuel-cladding interaction with the formation of U–Cs–Zr compounds was observed for burnup >40 GWd/t.

Modeling Chemical Interactions in the Hydrated Layers of Nuclear Waste Glasses.

The distribution of glass constituents within the hydrated layer is investigated through a mathematical representation of the fundamental processes that influence mass transport. A brief description of the models that are developed is presented. The resulting equations are implemented in a computer code named GELOH. The importance of the processes of network hydration and layer growth, alteration product formation, and dissolution is studied through the use of GELOH. The results of these studies are presented and interpreted in the light of the experimental evidence available on the aqueous corrosion of nuclear waste glass.

A criterion for selecting leach test specimen sizes

When leaching is controlled by a diffusion process, leach test results are particularly simple to interpret when test specimens approximate semi-infinite media. For spherical and cylindrical leach test samples, a criterion relating the test duration T, the specimen radius R, and the effective bulk diffusion coefficient D, to the desired degree of concurrence to the semi-infinite geometry behavior P, is shown to be given by: 1 - P = ..sqrt pi..DT/R. From the proposed criterion, it is concluded that, for glass waste forms, the semi-infinite geometry approximation is met by most test samples except possibly for finely crushed material.

The Functional Dependence of Leaching on The Surface Area-To-Solution Volume Ratio

The effects of the surface area-to-solution volume ratio on waste glass leach rates are investigated from a theoretical point of view. Simple leach models are discussed first. Correlation variables to interpret the results of similar leaching experiments performed at different values of the surface area-to-solution volume ratio are obtained for static leach testing. For dynamic leaching conditions, the source term required for risk assessment is derived and its dependence on the leachant flow rate and leach specimen surface area is discussed. The findings are upheld by a more complex leach model, the mathematical formulation of which has been implemented in a computer code named LIX. When tested against actual PNL 76-68 glass leaching data, LIX shows excellent capabilities in reproducing the experimental evidence, in particular the effects of the surface area-to-solution volume ratio.

Evaluating the Collective Radiation Dose to Workers from the U.S. Once-Through Nuclear Fuel Cycle

The Electric Power Research Institute (EPRI) is sponsoring the development of tools to support long-term strategic planning for research, development, and demonstration and for evaluation of advanced nuclear fuel cycles (NFCs). The EPRI-sponsored work under way at Vanderbilt University (VU) is developing a new, comparative risk assessment tool for NFCs. In the course of conducting a demonstration application of the assessment tool, it was observed that the relative contribution of NFC operations to radiological worker impacts estimated by the assessment tool was substantially different from widely used historical data and conventional wisdom. This paper analyzes these differences by first describing the NFC and the nature of radiological worker impacts. Then, the assessment tool developed by VU is described, along with assessment results; historical data relevant to radiological worker impacts are then summarized, and key differences between assessment results and previously reported impacts are identified. This comparison is followed by an analysis of the major factors causing the differences and an assessment of their validity. Finally, several implications of the differences are discussed.

Corrosion Performance Ranking of Zircaloy-2 for BWR Applications

The waterside nodular corrosion of Zircaloy cladding in boiling water reactors (BWRs) remains a primary concern in fuel performance because of (1) recurring fuel failures in several BWRs and (2) the current trend in the nuclear industry towards higher fuel burnups. The failures might be avoided if there were a reliable way of sorting claddings, so that only cladding that is highly resistant to corrosion would be used. In-reactor corrosion testing of cladding tubes is expensive and time-consuming, and consequently, there exists a strong incentive to develop a short-term, out-of-pile corrosion test that is able to predict cladding corrosion behavior for BWR applications. It is shown in this paper that a high-pressure steam autoclave test performed at 520°C for 24 h is capable of classifying the in-reactor nodular corrosion properties of the fuel cladding. The effects of test temperature, sample surface treatment, and oxygen and hydrogen content in the steam on corrosion performance are documented.

BWR pipe crack control using hydrogen water chemistry: Status report on Dresden-2 program

Abstract One of the proposed remedies for intergranular stress corrosion cracking of stainless steel piping in BWRs is an alternative water chemistry called hydrogen water chemistry (H 2 WC) that involves suppression of reactor water dissolved oxygen to ≤ 20 ppb via hydrogen injection to the feedwater in conjunction with control of conductivity to ≤ 0.3 μ mho/cm. A long-term verification program, over two or three 18 month fuel cycles, was started at Commonwealth Edisons Dresden-2 reactor in April 1983 (Cycle 9). This paper describes the results of the water chemistry changes, structural material and fuel evaluations, and plant radiation level changes during Cycle 9, which ended in October 1984. To date the results of the verification program are very encouraging. They indicate that the alternative water chemistry, based on hydrogen additions to the feedwater to suppress oxygen and low conductivity, can be maintained in a large operating BWR, and that it does mitigate IGSCC in stainless steel recirculation piping. Monitoring of fuel and plant materials will continue in Dresden-2 at least through Cycle 10 to confirm the absence of any unusual side effects of this remedy for IGSCC.

Effects of Fuel Relocation for Transport Casks

Spent nuclear fuel transported in large casks must remain subcritical in all credible configurations for normal operation and hypothetical accident conditions. The effects on spent nuclear fuel reactivity from “worst-case” accident scenarios were surveyed in NUREG/CR-6835, “Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks.” The survey used scenarios that were postulated to provide theoretical upper limits for reactivity effects of fuel relocation, although they were described as going “beyond credible conditions.” These scenarios involved physical changes either to fuel assembly rod arrays or to collections of fuel pellets with the fuel skeleton removed. To provide more credible estimates of the probability and maximum reactivity changes, a process is presented that deconstructs each scenario into a set of subscenarios and identifies the physical phenomena required to create the subscenario. The boundary between credible but unlikely scenarios and incredible scenarios is more easily discernible with this process. For marginally credible worst-case scenarios, it is concluded that the maximum reasonable reactivity increase either is less than the mandated administrative nuclear criticality safety margin for scenarios involving physical changes to fuel assembly rod arrays or is a substantial reactivity decrease for scenarios involving collections of fuel pellets. A cask designer could apply scenario deconstruction to evaluate the physical limits that apply to a particular transportation cask, and perform calculations specific to a particular cask design to show that criticality safety requirements are met.