[en] Post-dryout (PDO) temperatures and drypatch spreading have been measured in a 3-rod segmented bundle cooled by Freon-12. The bundle was tested both vertically and horizontally at mass fluxes ranging from 1 to 4 Mg.m^-2.s^-1. It was found that rod spacers 25 cm upstream of the end of the heated length caused considerable local CHF enhancement immediately downstream, resulting in a reduction of axial drypatch spreading. The fully developed PDO temperature data were successfully correlated using a high pressure water correlation. A correction term to account for partially developed PDO conditions was developed. This correction term was found to be applicable to high pressure water data from tubes and 37-rod bundle. The overall equation fitted the vertical Freon-12 PDO termperature data with rms error of 6.5% and the horizontal data with an rms error of11.9%. (auth)

[en] AECL is currently demonstrating the use of pool-type reactors of up to 10 MW output to produce hot water at about 90 degrees Celsius. The initial focus for the development is the provision of a source of hot water for institutional and municipal heating networks. Ongoing developments are designed to broaden the applications to electricity generation and industrial processes such as desalination and agricultural needs. The reactor concept is based on the Slowpoke-2 research reactor, eight of which are successfully operating in Canada and abroad. The primary-circuit flow is driven by natural convection, with the heated water, produced by the reactor core near the bottom of the pool, being ducted to low-pressure-drop heat exchangers in the upper part of the pool. As the pool volume is relatively large, the fluid transit time around the circuit is long, ensuring that the reactor response to all normal transients is extremely slow. To investigate thermalhydraulics aspects of the reactor design, including its behaviour underextreme conditions, an electrically heated, natural-convection loop was designed and constructed. The core of the loop consists of a rod bundle that is a precise reproduction of one quarter of the core of the 2-MW SLOWPOKE Demonstration Reactor presently being tested at the Whiteshell Nuclear Research Establishment. With this loop, measurements of the distribution of pressure, temperature, velocity and subcooled void have been made in the simulated core, via a variety of intrusive and non-intrusive techniques. In addition, both the single- and two-phase behaviour of the system have been studied. This paper gives examples of the various in-core measurements made and also makes comparisons between the measured system behaviour and that predicted by the various steady-state and transient computer codes

[en] Critical heat flux measurements have been made in a segmented 3-rod test section cooled by Freon-12. Three test section orientations were used: vertical, inclined at 11 deg to the vertical, and horizontal. It was found that at flows of less than 2.5 Mg.m^-2.s^-1 the transverse gravity force on the inclined and horizontal orientations reduced the magnitude of the critical heat flux and also changed the location of initial dryout when compared to the vertical data. To account for the effect of orientation during correlation of the data, the Reynolds number was modified to include a transverse gravity term. The minimum standard deviation for the data from the three orientations combined was 3.4 percent and less than 3.7 percent for the three orientations separately. (author)

[en] The authors are currently studying the feasibility of pool type reactors, for heating commercial buildings. To investigate thermal hydraulic aspects of reactor design, including the stability under extreme conditions, an electrically heated natural circulation loop was designed and constructed. Experimental tests with this loop show that it is stable under all single phase conditions, an expected result. In tests with two phases present in the circuit, different types of behaviour were observed depending on the experimental conditions. These were either stable flow, non-divergent oscillating flow or divergent oscillating flow. All of these type of behaviour were successfully simulated using the newly developed SPORTS stability computer code

[en] A natural circulation loop, or test rig, was built at the Chalk River Nuclear Laboratories for the investigation of the thermalhydraulic aspects of natural circulation at atmospheric pressure and corresponding subcooled or saturation temperatures. The test rig is a scale model of a small nuclear reactor intended for heating applications. A real-time, dynamic simulation of this test rig, code-named RIGSIM, has been developed and implemented on a hybrid computer. Step-transient experiments were carried out in the test rig to obtain the data for validating RIGSIM. Two series of tests were performed. In the first, a step change input, positive or negative, was applied to the test-section power; while in the second, the coolant-inlet temperature on the secondary side of the heat exchanger was stepped up or down. Results from both series of experiments show excellent agreement between RIGSIM predictions and experimental data, thus validating the RIGSIM simulation

[en] Detailed measurement of axial pressure profiles is often required to optimize the design of nuclear fuel assemblies. The classic method of using drilled pressure taps at predetermined locations in the test section has limited practicality and resolution. In addition, it cannot be used to test actual assemblies whose integrity must be preserved. This paper describes an alternative technique that can be applied to complex geometries. The method includes the use of a slender tube with a static hole near one end, which is introduced through slider seals into the test piece and moved to the desired locations. When signal-averaging is applied to the measurement, the technique is powerful in resolving pressure losses due to small subcomponents, such as fuel-element spacers. This technique was applied to the fuel design for a new research reactor, allowing the design to be optimized with confidence from the standpoint of pressure loss. (author)

[en] AECL is studying an advanced CANDU reactor concept, with supercritical steam as coolant. The coolant, being a high density gas, at a pressure above 22 MPa and temperatures above 370 deg C, does not encounter the two-phase region with its associated fuel-dryout and flow-instability problems. Increased coolant temperature leads directly to increased plant thermodynamic efficiency, thereby reducing unit energy cost through reduced specific capital cost and reduced fueling cost. The reduced coolant in-core density leads to sufficiently reduced void reactivity, so that light water becomes a coolant option. The use of supercritical water coolant also opens up the possibility of enhanced safety with a natural circulation primary flow, taking advantage of the gas expansion coefficient. To preserve neutron economy, especially at high coolant temperatures, a fuel channel that is currently being developed has a pressure tube that is thermally insulated from high-temperature coolant and is in contact with the cold heavy-water moderator. Two stages of development of a supercritical-cooled CANDU reactor were identified. The first uses conventional or near-conventional zirconium-alloy fuel cladding with coolant core-mean temperatures near 400 deg C, and the second uses advanced high-temperature fuel cladding at coolant core-mean temperatures near 500 deg C. A first-stage cost reduction of 20% from the CANDU 6 design is estimated as a result of improved thermodynamic efficiency. A large change in coolant density across the core leads to a factor 3 or 4 reduction in heavy-water inventory and a corresponding reduction in coolant void reactivity. The latter leads to improved fuel burnup and reduced demands on the safety shutdown systems. (author)

[en] Critical Heat Flux (CHF) measurements have been made in three subchannel shaped test sections cooled by Freon-12. The cross sections of the test sections were : round, triangular and rectangular ones. At a coolant quality of more than 10 % the effect of subchannel shape on CHF was not significant. The effect was significant, however, in the subcooled region where CHF was up to 38 % below that predicted by the W-3 tube correlation. (author)

[en] Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310^oC, and exits at ∼410^oC. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants (∼300 - 500 MWe). The steam cycle and coolant conditions are proposed to be the same as CANDU-X Mark I. The major difference between the reactors is that natural convection would be used to circulate the primary coolant around the heat transport system. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the critical point of water because of the large increases in heat capacity and thermal expansion coefficient across the core. The third concept, CANDUal-X, is a dual cycle concept, with core conditions similar to the Mark 1 and NC. In this concept, coolant leaving the core is first expanded through a VHP turbine in a direct cycle. Employing a dual steam cycle avoids a high-pressure steam generator. The conditions of the core and the VHP expansion can be designed such that the exhaust from the turbine is used as the heat source for an indirect cycle; that is, the secondary side can be equivalent to that presently employed in conventional CANDU plants. An advantage of this concept over conventional direct cycle nuclear plants is that only one relatively small turbine is exposed to radioactive coolant, and it is located within containment. In summary, the reactors described above represent concepts that evolve logically from the current CANDU designs to higher efficiency, with only modest extensions of current technology. This paper presents a technical overview of the different conceptual designs, as well as a brief discussion of the enabling technologies that are common to each, which is the focus of current R and D. (author)

[en] A natural circulation loop, or test rig, was built at the Chalk River Nuclear Laboratories for the investigation of the thermalhydraulic aspects of natural circulation at atmospheric pressure and corresponding subcooled or saturation temperatures. The test rig is a scale model of a small nuclear reactor intended for heating applications. A real-time, dynamic simulation of this test rig, code-named RIGSIM, has been developed and implemented on a hybrid computer. Step-transient experiments were carried out in the test rig to obtain the data for validating RIGSIM. Two series of tests were performed. In the first, a step change input, positive or negative, was applied to the test-section power; while in the second, the coolant-inlet temperature on the secondary side of the heat exchanger was stepped up or down. Results from both series of experiments show excellent agreement between RIGSIM predictions and experimental data, thus validating the RIGSIM simulation