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A reactor core overheats due to decay heat generated in the fuel when an effective cooling medium is unavailable, such as in a loss-of-coolant accident combined with a loss of emergency core coolant. If the heat generated is not effectively dissipated, then at extreme temperatures, the structural strength of the bundle assembly may deteriorate, leading to slumping of fuel elements onto the inner wall of the pressure tube. It is essential to examine the temperature behavior of the channel containing fuel pins in a disassembled state in order to comprehend the impact of further thermally induced deformations in the channel during postulated accident conditions. Capturing the temperature of channel components at each circumferential position from experiments is extremely difficult; thus, a modeling tool is necessary to obtain a thorough circumferential temperature profile. This paper presents a numerical study that aims to study the temperature distributions in a 1-m-long pressurized heavy water reactor (PHWR) channel containing a disassembled fuel bundle. The channel geometry and the boundary conditions implemented were obtained from the experiment. A temperature profile for each channel element at every circumferential and axial location was obtained. A thorough comparison of the predicted and the reported experimental values was performed, and it was found that the predicted temperature behavior of the channel was consistent with the experimental data. Further simulations with different fuel element configurations and decay powers may be carried out; in addition, the results obtained may be used for coupled thermal-mechanical and thermal-mechanical-chemical simulations. [ABSTRACT FROM AUTHOR]

The behaviour of the channel under postulated large break LOCA scenario had been a prime safety concern. The radiative heat transfer is predominant in a channel when the convective cooling environment is marred. The estimation of temperature distribution in the fuel pins at elevated temperature is essential from the point of view of hydrogen gas generation and release of fission products. In this paper, the thermal characteristics of a channel for Indian PHWR under critical break failure is studied using experimental and numerical techniques. The experiment is carried out on an Indian PHWR having a fuel bundle of 37-fuel elements. The temperature profiles for different parts of the simulated channel comprising of fuel pins, PT and CT are obtained under steady condition. The numerical analysis is also performed using ANSYS Fluent 19.0. From the study, it is found that there is a significant radial temperature gradient in the fuel bundle from the center ring to the outer ring. Also, no significant circumferential temperature gradient is observed in the fuel bundle, PT and CT.

Radiative heat transfer between the different components of the nuclear reactor channel (fuel pins, pressure tube (PT) and calandria tube (CT)) is the predominant mechanism of heat transfer in a channel when coolant flow is ceased. In this paper, the thermal behavior of the channel of Indian PHWR that has thirty-seven fuel elements is analyzed for a postulated accident such as loss of coolant accident (LOCA) coupled with the failure of the emergency core cooling system. (ECCS). The analysis has been performed using experimental, numerical and analytical techniques. The experiment is performed under the pseudo-steady-state condition, and the temperature profiles for the different parts of a simulated channel are obtained. The numerical analysis of the above problem is carried out using the ANSYS® Fluent 19.0. The code calculation is compared with the exact solution which is obtained using the radiosity network method. There is a reasonably good agreement between these three types of analysis. The results from these analyzes show that the temperature gradient (along the radial direction) is developed in the fuel bundle simulator (FBS) from the center fuel element to the outer ring fuel element. Also, an insignificant circumferential temperature gradient is observed in the FBS, PT and CT. [ABSTRACT FROM AUTHOR]

• Experiments were carried out in a non-oxidizing environment by varying the eccentricity of PT. • With the exception of the center pin simulator, a noticeable temperature gradient was established for an eccentric PT in the CT, PT, and bundle simulators. • The rate of heat dissipation from the bottom pin simulator was significantly higher than that from the top simulator for an eccentric PT. • For a PT eccentricity of 4 mm and 8 mm, the temperature at the bottom circumference of PT was lower than the upper circumference. • The lower circumference of the CT is at a higher temperature than the top circumference in the case of the eccentric PT. During a Loss-of-Coolant Accident (LOCA) combined with Loss-of-Emergency Core Coolant (LOECC), the convective cooling in the channel is degraded, resulting in a rise in bundle temperature due to heat decay despite the reactor being forced to shut down. The fuel bundle radiates heat to the Pressure Tube (PT), causing its thermomechanical properties to deteriorate at certain high temperatures and, depending on the channel pressure; PT may either sag or balloon. The assessment of the influence of eccentric PT on temperature distribution in the channel is critical for reactor safety. In the current study, various sets of experiments were carried out in a non-oxidizing environment by varying the eccentricity of PT, and for each of the experiments, the Joule heating of the bundle simulator was performed. A temperature profile was obtained for the bundle simulator, PT and Calandria Tube (CT). The results showed that in the channel, with the exception of the concentric arrangement, there was substantial temperature variation for all PT arrangements. [ABSTRACT FROM AUTHOR]

Under the accident condition, the loss of the continuous flow of coolant through the channel and the associated oxidative clad-steam chemical reaction causes an increase in the temperature of the fuel assemblies. If the moderator subcooling margin is insufficient, the channel may overheat to the point where the bundle assembly is jeopardized, and the fuel elements collapse onto the bottom of the Pressure Tube (PT). The present study aims to perform the numerical modeling of a severely degraded channel under steam flow condition from the thermal point of view. The configuration and the parameters employed in the study have been taken from the experiment. The energy equation incorporates all types of heat transfer, as well as secondary heat generated during the clad-steam oxidation reaction. The axial and circumferential temperature distribution has been obtained for the cladding, PT and Calandria Tube (CT), and compared with the experimental data. The predicted results are found to be in line with the experimental results, with the exception of the temperature excursion in the Fuel Bundle simulator (FBS) downstream of the coolant. • Numerical modeling of a severely degraded channel under steam flow condition has been performed. • A detailed comparison has been made between the predicted and experimental temperature values. • The predicted results are found to be consistent with the experimental results, except for a temperature excursion in the bundle simulator downstream of the coolant. [ABSTRACT FROM AUTHOR]

• An experimental system was created to simulate the late phase of the accident scene. • The experimental system also simulated the sagging deformation of the PT. • A numerical simulation was performed for the channel with the PT-CT contact configuration under an oxidizing environment. • Significant axial and circumferential temperature gradients were established in the channel compared to that obtained under a non-oxidizing environment for a similar test temperature. The loss of primary as well as emergency coolant flow causes the convective cooling of the fuel bundle to degrade, raising the temperature of the fuel channel. Secondary heat released by the Zr/steam chemical reaction heats the channel even more intensely. The present work examines the thermal behavior of a heated channel in an oxidizing environment. An experimental system was created to simulate the late phase of the accident scene. The Pressure Tube (PT) in the channel was configured to contact the Calandria Tube (CT) at the bottom, simulating sagging deformation. The simulation was performed using ANSYS Fluent software. The modeling of the effect of the Zr/steam chemical reaction was done by adding a heat source to the cladding. A comparison was made between the obtained results and the available results of the same channel configuration in a non-oxidizing environment. It was found that significant axial, as well as circumferential temperature gradients were developed in the channel compared to that obtained under a non-oxidizing environment. [ABSTRACT FROM AUTHOR]

• The Thermal characteristics of the IPHWR specific fuel bundle with disassembled fuel pins configuration is analyzed. • The configuration considers the collapse of 37 fuel pins of a bundle from the endplate. • A notable circumferential temperature gradient is observed for outer fuel pin simulators that are in contact with PT. • The significant circumferential temperature gradient is developed in the PT due to the slumping of fuel pins. • Due to PT-CT contact, a sudden temperature drop at the bottom nodes of the PT is observed. The thermal behavior of reactor channels is of prime concern for a postulated scenario of LOCA along with the un-availability of ECCS as the integrity of the fuel bundle may be challenged and slumping of fuel pins is expected. The present study is aimed to analyze the thermal characteristics of a large capacity Indian PHWR specific fuel bundle with disassembled fuel pins configuration. The configuration considers the collapse of 37-fuel pins of a bundle from the endplate and rearranges at the bottom surface of the PT. The paper further investigates the influence of PT-CT contact on the channel. The results show that a significant circumferential temperature gradient is developed in the PT due to the slumping of fuel pins. The maximum temperature is obtained at the bottom sector of the PT, which is in contact with the fuel pin. However, PT-CT contact causes a sudden temperature drop at the bottom nodes of the PT. [ABSTRACT FROM AUTHOR]

Highlights • The pseudo state temperature of 900 °C−1000 °C was obtained in the fuel bundle for 1% decay power. • Insignificant circumferential temperature distribution was found in the fuel pins. • 85% Decay heat was removed by moderator during large break LOCA. Abstract Postulated accidents like large break LOCA leads to expulsion of coolant in the primary heat transport system thus voiding of reactor core initially. However, the reactor is shut down at the onset of LOCA along with coolant injection from ECCS to remove the decay heat of 2–3% of nominal power. Further postulation of failure of ECCS leads to rapid increase in the temperature of fuel pins and the coolant channel as well. The moderator present around coolant channels limits the rise in temperature. The assessment of temperature distribution in the fuel pins of the bundles of a channel under high temperature is quite important from hydrogen generation and fission product release point of view. A pseudo steady state experiment has been carried out to obtain the temperature distribution of different components of a simulated coolant channel for large capacity Indian pressurized heavy water reactor. The experiment simulates a postulated LOCA with Loss of ECCS scenario for the coolant channel. The experimental results shows that for 1% decay power the simulated fuel pins attained a maximum pseudo steady state temperature of 900 °C–1000 °C, thus establishing moderator as a heat sink. An insignificant circumferential temperature variation in the channel components is observed which indicates weak effect on natural convective heat transfer within coolant channel. Estimation shows that around 85 percent of total decay heat provided through electrical power is removed by the moderator. [ABSTRACT FROM AUTHOR]

• The PT/CT contact develops a significant circumferential temperature variation in both PT and CT. • The maximum temperature is obtained at the top and bottom surface of PT and CT respectively. • Significant circumferential temperature distribution is also found in the fuel bundle expect in the fuel pin of ring-1. • Moderator acts as an emergency heat sink during large break LOCA as it takes away around 84 percent decay heat. An experimental investigation has been performed to simulate a scenario of a beyond design basis accident like LOCA along with the un-availability of ECCS. The investigation is aimed to study the thermal behaviour of a single channel of PHWR during sagging deformation. The sagging deformation of PT was simulated by making a direct physical contact between PT and CT. The fuel channel assembly of a length of 1.4 m was immersed in water which simulates Moderator. The decay heat liberated after reactor shutdown was simulated by Joule heating of the 37-fuel pin elements. The steady state circumferential temperature distribution in the 37-fuel pin bundle simulator, PT and CT at different axial position has been obtained. It was found that there is a significant circumferential temperature gradient in the fuel channel. The maximum and minimum temperature was obtained respectively at top and bottom nodes of PT. However, an opposite trend was observed in CT. The maximum temperature difference of 185.7 °C and 10.2 °C has been found between top and bottom surfaces of PT and CT respectively. A similar behaviour to that of PT was discerned in the 37-pin fuel element. However, fuel pin of ring-1 has an insignificant effect of PT-CT contact. It was observed that around 84 percent of simulated decay heat is transferred to the simulated moderator. [ABSTRACT FROM AUTHOR]