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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.

The postulated severe core damage accidents in PHWR can get initiated from limited core damage accidents incase moderator cooling is lost. Excessive sagging of the exposed fuel channels from moderator boil-off leads to channel disassembly. Steam reacts with fresh zircaloy at higher temperatures and generates hydrogen to a large extent. The present work investigates the effect of ingression of steam and associated hydrogen generation on a simulated disassembled channel for 220 MW(e) PHWR under postulated severe core damage accident conditions. The assessment of steam ingression in the disassembled channel at various inclination angles is carried out. The study shows that the inclination angle highly influences the ingression of steam, fuel channel temperature and hydrogen generation. The very low steam ingression for horizontal position of the channel was found to limit the Zr oxidation, while at higher inclinations angle, sufficient amount of steam ingression leads to higher hydrogen generation.

Quenching studies of ballooned fuel pins have indicated enhancement of coolability for flow blockages ranging upto 90% with blockage extension of 6% (20 cm blockage length) under typical Emergency Core Cooling System (ECCS) injection rates. Similar enhancement is also observed for flow blockage of 45% with higher blockage extension of 60% and lower injection rates. An experimental setup is developed to assess the coolability under high flow blockage (upto 80% of the flow area) and longer ballooned length extensions (up to 60% or 600 mm). The setup employs 5 X 5 matrix of indirectly heated, pre-fabricated ballooned fuel pin simulator (FPS) surrounded with 20 heated and ballooned FPS which are further surrounded with 12 dummy FPS. The objective of this experiment is to study the effect of water injection rate on the quenching behaviour of large scale ballooned heated pins simulating early phase of severe accident. Bottom re-flood condition is considered for the study. The water injection rates (0.11–0.45 g/s per unit length per FPS) are kept lower than the typical PWR specific SAMG injection flow rates to assess minimum flow rate requirement. The FPS is observed to be coolable only when the injection rates are higher than a certain value. Higher quenching rate is observed in the region towards the entry of the ballooned length as compared to the region towards the exit of the ballooned length. Conduction controlled rewetting is found to be dominant for the entire range of injection rates considered for the experiments. Flow rates (0.11–0.45 g/s per unit length per FPS) are found to successfully quench the bundles. However, the FPS temperatures exceed the oxidation run-away threshold temperature for 10–25 g/s injection flow rates (0.11–0.275 g/s per unit length per FPS).

A postulated scenario of station blackout (SBO) or loss of coolant accidents (LOCA) with the failure of all safety systems can lead to the unmitigated severe accident in pressurized heavy water reactors (PHWRs). As water is present as moderator in the Calandria vessel participates in decay heat removal hence it undergoes boil off. This initiates the slow uncovering of the top channels. Continuous heating of exposed channels may lead to disassembly of the channels. Under such circumstances, the steam can ingress inside the disassembled channels and react with unreacted zirconium and generate more hydrogen. The present paper aims to assess the thermal behavior, steam ingression inside the channel and associated hydrogen generation of standard 220 MWe Indian PHWR under the hypothesis of postulated severe accidents with large break LOCA and loss of ECCS as an initiating event. Experimental and numerical study has been carried out for a one-meter channel length for 2% of decay power level. Transient CFD simulation is also performed for a better understanding of the thermal behavior of the channel. Throughout the transient, the fuel bundle of the channel is heated maximum up to the temperature 891 °C, and generation of hydrogen is observed beyond maximum temperature of the fuel bundle of 628 °C. The average hydrogen generation rate is found to be 0.01822 g/s. The experiment shows that oxidation due to steam ingression is limited to the regions near the end of the channels. Radiation and convective steam cooling helps to limit the rise in the temperature of the fuel bundle. Ingression effect is noticed up to 260 mm from ends, beyond which limited steam is available.

The safety demonstration of nuclear power plant, even for very low probability severe accident scenarios, has been an important requirement. A postulated severe accident is one of the beyond design basis plant states which can have significant consequences resulting from the nuclear fuel degradation. Such accident scenarios call for detailed safety assessment and provision of design features and guidelines for the operating personnel to manage the accident. In the present work, structural integrity assessment of Calandria assembly of a typical 540 MWe Indian Pressurized Heavy Water Reactor has been carried out to assess the duration of in-Calandria retention of core debris/corium for a postulated severe core damage accident scenario. The premise of the study, methodology and analysis tools used to assess structural integrity and failure timeline are discussed. Sequentially coupled thermo-mechanical analysis has been carried out for the thermal and mechanical loads arising out of the postulated accident and structural response of Calandria assembly is evaluated. The Calandria undergoes significant creep/plasticity deformation after loss of cooling through Calandria vault water. Plastic instability, excessive inelastic strains and creep-stress rupture criteria are considered to assess the structural failure of Calandria assembly. It is observed that the Calandria is able to maintain the structural integrity for reasonable duration of more than two days without any management action. Thus, the robustness of the PHWR design is demonstrated. Realistic assessment of timeline of in-Calandria corium/debris retention under such severe core damage accident loading are vital for assessing the severe accident management guidelines.

The present investigation reports the rewetting phenomenon of a flat surface where the water front approaches the centre of a heated plate from all the edges. The water front further develops a water pool above the plate surface. The typical transient temperature response for the combination of flow and developing pool boiling has been investigated. The effect of several parameters, including initial surface temperature (300 °C to 600 °C), coolant flow rate (25 g/s to 63.3 g/s) and input wall heat flux (44 kW/m2 to 66 kW/m2) on rewetting patterns have been studied. It is observed that the rewetting of the flat plate is sensitive to initial surface temperature and coolant injection flow rate. Increase in initial surface temperature and decrease in coolant injection flow rate results in delayed rewetting whereas variation in input heat flux has a marginal effect on peak surface temperature and time delay for initiation of a rewetting phenomenon. The observed physical phenomena include movement of the waterfront from plate edges towards its centre, leading waterfront edge vaporisation, detachment of water droplets from leading edge, the formation of the water pool, temporary generation of vapour columns and mist formation within the developing pool. A film boiling correlation has been proposed for predicting the heat transfer coefficients for the rewetting of a flat plate. Mass flux parameter is introduced in the correlation to address the effect of different reflooding rate, which is not prevalent in the traditional film boiling correlations. The proposed correlation can predict the heat transfer coefficient in the range of ±15% of the experimental observation.

Quenching of reactor core has been studied numerically with the help of several numeric tools reported in literature. These tools have been developed to simulate core quenching under design basis conditions. Hence, most of these tools make use of classical homogeneous equilibrium model of two phase flow which are reported to perform better under high pressure conditions with relatively small temperature gradients. System codes such as RELAP5 make use of two fluid model. Such system codes can simulate core quenching under design basis injection rates (1 g/s per unit length of single fuel pin) prescribed under Emergency Operating Procedures (EOPs). However, because of possibilities of flow path breakage under Severe Accident conditions, the injection rate reaching the core is expected to be lesser than the typical injection rate used for Severe Accident Management Guidelines (SAMG). A numeric tool ‘Program for Degraded Reactor Core Reflood’ (PDRCR) has been developed to simulate quenching of partially degraded reactor core under low injection flow rates. PDRCR has specific modules to tackle issues such as water packing, high temperature gradients in the fuel and simulation of parallel flow paths to account for flow re-distribution in partially ballooned fuel pin condition. This paper presents validation results for PDRCR. The predictions of PDRCR for SEFLEX and DRCRE test facilities and comparison with the experimental results are presented. The predictions of PDRCR are found to be better than RELAP5 for low injection flow conditions.

The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases.

The Pressurised Heavy Water Reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Usually for all these designs the Emergency Operating Procedures (EOPs) are developed in support of detailed accident analysis, which gives an adequate coverage for design basis accidents. Currently the designers are making provisions [1& 2] in design to mitigate progression of accidents arising from multiple failure accidents like Large Break Loss of Coolant Accident with failure of Emergency Core Cooling System and failure of moderator as heat sink. Many designs of Large PHWRs have adopted the approach of symptom based EOPs to handle multiple failure events as currently practiced for Light Water Reactors (LWRs). Severe accident analysis is an important aspect which complements Severe Accident Management Guidelines (SAMG) development process. These analysis provide insight into the accident progression and basis to develop the SAMG. The order of uncertainty in modelling the phenomena is very high. Hence it is emphasized that different computational models be used so that an un-biased “insight” can be evolved which can be used for SAMG development. The paper discusses two categories of severe accident analyses for such large reactors for multiple failure transients involving a high pressure scenario (initiation event like SBO) and low pressure scenario (initiating event like LOCA). The insight evolved from these analysis is being discussed in the paper.