The containment of an Indian pressurized heavy water reactor (IPHWR) is the ultimate physical barrier, designed to enclose the reactor systems and prevent the release of airborne radioactive fission products during both normal operation and accident conditions. The containment consists of several compartments, one of which is the fueling machine vault (FMV), which houses the fueling machine and components of the primary heat transport (PHT) systems, such as headers and feeders to the core. The FMV is connected to the pressure relief chamber (PRC) and steam generator vault (SGV) to relieve pressure during a loss of coolant accident (LOCA). To protect the containment from the threat of hydrogen accumulation, large openings are required between the FMV and the SGV or PRC to enable pressure equalization and hydrogen dispersion. However, these openings prevent the formation of enclosed volumes for effective cooling of the FM-vault, resulting in the reduction in the reactor building (RB) cooling, which leads to a rise in temperatures within the containment, particularly in the pump room area and dome region during normal operation. The air cooling units (ACUs) in the reactor building remove heat generated by the equipment and piping within the containment. To increase the heat removal capacity of the ACUs, the water in the tube side of the heat exchangers has been replaced with chilled water, which is powered by Class-IV power supply during normal operation. In the event of a Class-IV power failure, this cooling would not be available. To ensure containment cooling under such conditions, two ACUs are supplied with active process water (APW), which operates on a Class-III power supply. The passive decay heat removal system (PDHRS), a first-of-a-kind (FOAK) system in the 700 MWe reactor, plays an important role in cooling the reactor core during a station black-out (SBO) event. During SBO, the loss of Class-IV and Class-III power supplies results in the complete loss of active heat removal mechanisms (reactor building coolers) from the containment atmosphere. During the SBO event, the reactor building (RB) gets isolated due to the closure of the ventilation fan dampers. Under such conditions, containment analysis shows that the pump room pressures rise to 12 g/cm2(g) in the early phase of the transient, which would initiate the crash cool down (CCD), making PDHRS ineffective. Therefore, to ensure the effectiveness of the PDHRS, the CCD set point has been raised to 30 g/cm2(g) of the pump room pressure. The containment system is provided to limit the release of radioactive fission products to the environment from the reactor core and reactor coolant system during and after accident conditions. For postulated accidents considered in the design basis, the estimated release shall be within acceptable limits. Design basis events for the design requirements of the containment, e.g. its pressure rating, leak tightness, and performance requirements of the ESFs are LBLOCA, MSLB, and LOCA with impairment of ECCS. All calculations and analyses have been performed using the in-house computer code PACSR-SI-2.0. Although PACSR-SI-2.0 has been validated for various postulated accident scenarios involving highly turbulent conditions and significant changes in mass and energy release into the containment, the changes during a Class-IV power failure are relatively smaller in magnitude. Therefore, it is necessary to validate the computer code for such scenarios. To this end, a partial Class-IV power failure test was conducted to assess the rise in pressure and temperature within the containment. The test results were compared with the pressure and temperature predictions made by the PACSR-SI-2.0 code. The containment analysis showed that the predicted values either matched or were slightly conservative compared to the experimental data. The paper covers all the above aspects of 700 MWe IPHWRs.

Containment system for 700 MWe IPHWRs

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发表日期：

2026年7月18日

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Sanjeev Kumar Sharma,
Manvendra Singh,
Ankush Yadav,
Vibha Hari,
Sameer Hajela

摘要

The containment of an Indian pressurized heavy water reactor (IPHWR) is the ultimate physical barrier, designed to enclose the reactor systems and prevent the release of airborne radioactive fission products during both normal operation and accident conditions. The containment consists of several compartments, one of which is the fueling machine vault (FMV), which houses the fueling machine and components of the primary heat transport (PHT) systems, such as headers and feeders to the core. The FMV is connected to the pressure relief chamber (PRC) and steam generator vault (SGV) to relieve pressure during a loss of coolant accident (LOCA). To protect the containment from the threat of hydrogen accumulation, large openings are required between the FMV and the SGV or PRC to enable pressure equalization and hydrogen dispersion. However, these openings prevent the formation of enclosed volumes for effective cooling of the FM-vault, resulting in the reduction in the reactor building (RB) cooling, which leads to a rise in temperatures within the containment, particularly in the pump room area and dome region during normal operation. The air cooling units (ACUs) in the reactor building remove heat generated by the equipment and piping within the containment. To increase the heat removal capacity of the ACUs, the water in the tube side of the heat exchangers has been replaced with chilled water, which is powered by Class-IV power supply during normal operation. In the event of a Class-IV power failure, this cooling would not be available. To ensure containment cooling under such conditions, two ACUs are supplied with active process water (APW), which operates on a Class-III power supply. The passive decay heat removal system (PDHRS), a first-of-a-kind (FOAK) system in the 700 MWe reactor, plays an important role in cooling the reactor core during a station black-out (SBO) event. During SBO, the loss of Class-IV and Class-III power supplies results in the complete loss of active heat removal mechanisms (reactor building coolers) from the containment atmosphere. During the SBO event, the reactor building (RB) gets isolated due to the closure of the ventilation fan dampers. Under such conditions, containment analysis shows that the pump room pressures rise to 12 g/cm²(g) in the early phase of the transient, which would initiate the crash cool down (CCD), making PDHRS ineffective. Therefore, to ensure the effectiveness of the PDHRS, the CCD set point has been raised to 30 g/cm²(g) of the pump room pressure. The containment system is provided to limit the release of radioactive fission products to the environment from the reactor core and reactor coolant system during and after accident conditions. For postulated accidents considered in the design basis, the estimated release shall be within acceptable limits. Design basis events for the design requirements of the containment, e.g. its pressure rating, leak tightness, and performance requirements of the ESFs are LBLOCA, MSLB, and LOCA with impairment of ECCS. All calculations and analyses have been performed using the in-house computer code PACSR-SI-2.0. Although PACSR-SI-2.0 has been validated for various postulated accident scenarios involving highly turbulent conditions and significant changes in mass and energy release into the containment, the changes during a Class-IV power failure are relatively smaller in magnitude. Therefore, it is necessary to validate the computer code for such scenarios. To this end, a partial Class-IV power failure test was conducted to assess the rise in pressure and temperature within the containment. The test results were compared with the pressure and temperature predictions made by the PACSR-SI-2.0 code. The containment analysis showed that the predicted values either matched or were slightly conservative compared to the experimental data. The paper covers all the above aspects of 700 MWe IPHWRs.