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Highlights • Sliding pressure sensors are developed for flow characterization in complex geometries. • 3-D pressure mappings help visualizing flow characteristics in the vicinity of grid spacers. • Dynamical analysis of the pressure signals allows identifying structures in the flow. • Technique able to study interference and correlation between complex flow structures. Abstract Investigating the flow behavior in fuel rod bundles has been an active research topic since many decades. Nowadays, despite the great advances in computational fluid dynamic techniques and resources, blind benchmarks have shown numerical results are still very dependent on schemes and closure models. Moreover, most experimental studies in flow mixing and flow structure identification were performed in simplified geometries and/or in a very limited test domain. In addition, reliable experimental data from mixing experiments in the vicinity of a spacer grid in rod bundles are almost non-existent, especially with high spatial and temporal resolution. Generating experimental data relevant for fuel bundle geometries is costly and generally limited to a small measurement region. In this work a novel non-intrusive technique is developed and tested in order to characterize the flow in a geometry resembling a fuel bundle by using of detailed pressure measurements. The measurement device makes use of electronic micromachined deformable membrane differential pressure sensors with fast dynamical response, allowing capturing local pressure fluctuations. Such differential pressure sensors are connected to a fix pressure tap and a movable pressure tap drilled in each of the rods. Each instrumented rod is free to move both axially and azimuthally allowing scanning the static pressure drop values at the surface of the rod. In addition, by analyzing the static pressure fluctuations it is possible to capture valuable information of turbulence phenomenon such as the turbulence kinetic intensity and the turbulence power spectrum in a wide spatial range, including regions within spacer grids. [ABSTRACT FROM AUTHOR]

Fuel bundle end power peaking in pressure tube heavy water reactors (PT-HWRs), particularly considering advanced reactor fuels (such as thorium-based fuels), was evaluated and a mitigation method was assessed. For bundles containing all the same fuel pellets, power is greatest at the bundle ends due to higher thermal neutron flux there. A method to achieve a flatter axial power profile along a bundle by downblending the fissile content in the end pellets with thorium dioxide was evaluated. The method was shown to be effective, and only a slight reduction in fresh fuel reactivity was observed. [ABSTRACT FROM AUTHOR]

For a large break Loss of Coolant Accident (LOCA) scenario with simultaneous failure of Emergence Core Cooling System (ECCS), the heat removal capability of coolant decreases, resulting in the complete voiding of the core of the reactor channel. In this scenario, radiation becomes dominating mode of heat transfer over the convection heat transfer from fuel bundle to Pressure Tube (PT). Due to this, the deformation of PT takes place either in the form of ballooning or sagging depending on prevailing internal pressure. The present paper aims to capture the circumferential temperature profile over the fuel bundle, PT and Calandria Tube (CT) undergoing sagging deformation of PT under a postulated LOCA with Loss of ECCS scenario. An experimental facility was designed and fabricated to simulate the sagging deformation of PT by changing the eccentricity “e” of PT with respect to the CT. The experiments were conducted at three different positions of PT i.e. e = 0, e = 4 mm and e = 8.5 mm at 1100 °C fuel bundle temperature. The experimental results showed that the eccentricity of PT strongly affected the circumferential temperature distribution over the fuel bundle, PT and CT. The variation of temperature along the circumference of fuel bundle and PT at e = 0 was insignificant, however, for other positions, this variation in temperature was quite significant. The bottom nodes of fuel bundle showed a decrease in temperature with an increase in eccentricity. Similar observations are also made for PT. On the other hand, the temperature at the bottom nodes of the CT increased for higher eccentricity between PT and CT. The experiments predict the temperature behavior for all the channel components for a decay power of 2% full power which is expected during such scenario. [ABSTRACT FROM AUTHOR]

The safety of the nuclear reactor depends on the heat transfer of the supercritical water in the fuel bundle. Investigations of effects of operating conditions on the heat transfer of the nuclear reactor are still lacking. In this study, the effects of the operating pressure, inlet temperature, heat flux, and mass flux on the heat transfer of upward supercritical water flow in the 64-element fuel bundle were studied numerically. The criteria of the onset of buoyancy effect on the heat transfer deterioration in circular tubes were also used to evaluate the buoyancy effect on the heat transfer in the fuel bundle. The results show that the wall temperature generally increases with the increase in the inlet temperature, heat flux, or the decrease in the mass flux. The buoyancy-affected zone mainly exists at the pseudocritical temperature region and it will disappear when the mass flux is above a certain value. [Display omitted] • The heat transfer of the supercritical water in the fuel bundle at different operation conditions was investigated. • Criteria of the onset of buoyancy effect on the heat transfer deterioration were evaluated. • The buoyancy-affected zone mainly exists at the region where the bulk temperature is close to the pseudocritical temperature. [ABSTRACT FROM AUTHOR]

Laminar flow of sodium in the entrance region of a seven pin nuclear fuel bundle with helically wound spacer wires is numerically investigated. Pin diameter, pin pitch, helical wire pitch and Reynolds number are varied as parameters. Evolutions of cross-stream velocity, friction factor and Nusselt number are critically studied. Nusselt number and friction factor exhibit periodic spatial oscillations in the entrance region, contrary to monotonic reduction observed in straight ducts. For identical Reynolds number, the friction factor exhibits a strong dependence on helical pitch of wire wrap where it increases as the pitch shortens. On the contrary, the Nusselt number does not exhibit such a strong dependence on helical pitch. A strong non-zero cross-stream velocity prevails in the fully developed region, contrary to that observed in straight channels. The mean value of non-dimensional cross-stream velocity is found to scale as πd / H , where, d is the pin diameter and H is the helical pitch. The friction factor and cross-stream velocity are relatively high in tight pin bundles. Based on parametric studies, a correlation is proposed for fully developed friction factor in pin bundle. [ABSTRACT FROM AUTHOR]

• The PT/CT contact leads to the high heat transfer to moderator. • The highest temperature reached in the PT for 4500 W/m2 heat input in fuel bundle was 1898 °C. • For 4500 W/m2 heat input rate the volume fraction of moderator vapour was maximum. • Upto 4500 W/m2 moderator acted as an efficient heat sink. For an Indian pressurised heavy water reactor, a numerical investigation was performed to study the boiling of moderator in a fully voided single nuclear channel during Pressure tube- calandria tube contact condition by applying variable heat flux. The setup included calandria tube, moderator, pressure tube, fuel bundle and an enclosure. It was observed that because of the contact between calandria tube and pressure tube, a notable variation in the circumferential temperature distribution of calandria tube and pressure tube were observed pointing out that contact condition drastically affects the temperatures of top and bottom portion of the tubes. It was also observed that upon increasing the heat flux, the top to bottom temperature variation decreases for pressure tube but increases for calandria tube. The phenomenon of boiling of moderator started when 1650 W/m2 heat flux was given to the system as input and the rate of boiling increased as the heat flux was increased which resulted in increasing moderator vapour fraction in the enclosure and decrease in average moderator density. The film thickness of the moderator vapour on the outer circumference of the calandria tube varied with time, following a cycle of increasing and decreasing thickness. The result showed that upto 4950 W/m2 heat flux, the moderator acted as an efficient heat sink. If the heat flux is increased further, the moderator can fail to act as a good heat sink and eventually lead to fuel channel failure. [ABSTRACT FROM AUTHOR]

Abstract: This paper describes in brief the need for the coolant channel replacement, the challenges faced, the methodology adopted to successfully overcome the shortcomings, and the improvements made in various areas that set a new standard in the global nuclear industry. These changes have streamlined the en-masse coolant channel replacement programme which was followed during the subsequent campaigns. [Copyright &y& Elsevier]

Abstract: During a loss of coolant accident in nuclear reactors, radiation heat transfer accounts for a significant amount of the total heat transfer in the fuel bundle. In case of heavy water moderator nuclear reactors, the decay heat of a fuel bundle enclosed in the pressure tube and outer concentric calandria tube can be transferred to the moderator. Radiation heat transfer plays a significant role in removal of decay heat from the fuel rods to the moderator, which is available outside the calandria tube. A single heater rod test facility is designed and fabricated as a part of preliminary investigations. The objective is to anticipate the capability of moderator to remove decay heat, from the reactor core, generated after shut down. The present paper focuses mainly on the role of moderator in removal of decay heat, for situation with air alone in the annular gap of pressure tube and calandria tube. It is seen that the naturally aspirated air is capable of removing the heat generated in the system compared to the standstill air or stagnant water situations. It is also seen that the flowing moderator is capable of removing a greater fraction of heat generated by the heater rod compared to a stagnant pool of boiling moderator. [ABSTRACT FROM AUTHOR]

• Simulations of the axial flow in CANDU fuel bundles were conducted. • Motion dependent fluid forces were calculated. • A numerical model that simulates fuel bundle vibrations was developed. • The model prediction compares well with the experimental results. Fuel bundles in CANDU (Canada Deuterium Uranium) reactors are subject to axial flow-induced vibrations. This could result in fretting wear in the inner surface of the pressure tube or in the integrity of the bundles being compromised. Various fluid excitations can affect the dynamic response of the fuel bundle. One of these excitations is the fluid forces produced by the motion of the fuel rods. These forces can be expressed in terms of inertia, stiffness, and damping components. A numerical approach was proposed to simulate these forces in a flexible fuel kernel. These forces were expressed in terms of coefficients and phases. In addition, an analytical model was developed utilizing the force coefficients to predict the dynamic response of the fuel bundle. The dynamic response was compared with the available experimental data. The numerical model developed in this work presents a step towards a realistic framework to simulate the dynamic response of CANDU fuel bundles. [ABSTRACT FROM AUTHOR]