[en] The kinetics of the α → β phase change in Zr-2.5 percent Nb pressure-tube material at constant temperature have been studied. The volume-fraction change of the β phase due to diffusion in an infinite α-phase matrix was considered, and a mathematical model with a numerical solution was developed to predict the transient spherical growth of the β-phase region. This model has been applied to Zr-2.5 wt percent Nb, and the calculated results compared to experiment

[en] A computer code called SLUDGE has been developed that can be used to model secondary-side steam generator (SG) fouling as a function of time and operating conditions while still providing details of the deposit distribution throughout the SG. Tube deposits are predicted to grow at average rates of 4.4 and 1.1 μm/a on the hot side and cold side of the bundle, respectively. This is in good agreement with available plant data. SLUDGE under-predicts the magnitude of the tubesheet deposit, which implies that mechanisms in addition to gravitational settling may be responsible. One such mechanism, centrifugal separation of particles from the fluid streamlines, increased deposition towards the periphery of the tubesheet, but did not predict a realistic sludge pile. Comparison between the deposits on the horizontal surfaces of the tube support plates and the tubesheet suggests that the rate of particle removal at the tubesheet may have been overestimated. SLUDGE predicts a kidney-bean shaped sludge pile on the tubesheet when the re-entrainment rate is set equal to zero. (author). 17 refs., 1 tab., 6 figs

[en] At the time of the conference, the ASSERT code was being developed to address the three-dimensional computation of flow and phase distribution and fuel-element surface temperatures on a local or subchannel basis within CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) throughout the fuel channel. This paper briefly describes the ASSERT code and its CHF assessment methodology, and several recent validation exercises concerning applications of the code to simulations relevant to CANDU-PHWR fuel channel conditions. Results are presented of a preliminary ASSERT analysis of the Stern Laboratory full-scale water critical heat flux tests. The initial data from the nominal (uncrept) pressure tube tests for steady-state onset-of dryout conditions was applied to ASSERT. An advanced (pre-experiment) prediction of the Stern Laboratory tests in a crept pressure tube was completed using ASSERT. The purpose of the study was to give the relative reduction in initial fuel channel dryout power for a crept pressure tube compared to the nominal pressure tube for the same boundary conditions. This paper discusses recent ASSERT code validation of two CHF experiments with full-scale vertical flow bundles with Freon coolant; these experiments discussed the effect of fuel bundle radial heat flux distribution (RFD) on CHF. The purpose of the study was to confirm the ability of the ASSERT code to predict the correct parametric trends in dryout power for various RFD profiles. Some past ASSERT validation studies relevant to CANDU-PHWR fuel channel modelling are briefly reviewed. The status of, and future directions for, ASSERT code development are summarized. 40 refs., 9 figs., 1 tab

[en] Atomic Energy of Canada Limited is studying the disposal of nuclear-fuel-waste in a vault in plutonic rock. An important part of this study is to develop models to assess the rate of escape of waste components to the biosphere. In these models time-dependent changes in the system are important. One such change is the increase in temperature of the rock mass immediately surrounding the vault, and in the far-field. Differential-thermal expansion between rock constituents (crystals), and the expansion of trapped water, will produce microcracking and, hence, possibly changes in intact-rock elastic modulus and permeability. This paper describes the model devised to estimate the extent of microcracking in rock surrounding a vault, and the anticipated changes in rock elastic modulus and permeability. The model is applied to a possible vault, located at various depths in plutonic rock typical of the Canadian Shield. It estimates the dimensions of an envelope outside of which the influence of the vault on microcracking will be negligible

[en] Atomic Energy of Canada Limited is studying the possible disposal of nuclear-fuel waste in a vault in plutonic rock. An important part of this work is to develop models to assess the rate of escape of waste components to the biosphere. Clearly, these models must account for time-dependent changes in the system. One such change is that heat from the waste will raise the temperature of the rock surrounding the vault. This will produce microcracking at the interfaces between rock constituents, due to differential thermal expansion plus some contribution from the expansion of trapped water. As the temperature of the rock varies with time and position relative to the vault, so will the microcracking. This paper describes an approach to estimating the possible extent of thermal-induced microcracking around a vault. From the estimated development of the microcrack population, possible changes in rock stiffness and permeability (for intact rock) are calculated, as a function of time and position

[en] The deposition of solid particles from two-phase flow on surfaces is a complicated problem. It is further complicated in steam generator analysis by the need to simultaneously model the heat transfer and its response to deposition. This paper describes a three-dimensional transient simulation of magnetite particulate fouling, and the accumulation of particles on the heat transfer surfaces in a nuclear steam generators