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We report the mass-selected slow photoelectron spectra of three reactive organophosphorus species, PCH2, and the two isomers, methylenephosphine or phosphaethylene, HPCH2 and methylphosphinidine, P–CH3.

Empirical correlations are developed for rewetting velocity and maximum heat transfer coefficient during rewetting phase of single hot vertical Fuel Pin Simulator (FPS) by using radial jet impingement and falling film. Emergency Core Cooling System (ECCS) has been designed for Advance Heavy water Reactor (AHWR) to rewet the hot fuel pin under the loss of coolant accident. Coolant injection takes place from a water rod which is located at the center of the fuel bundle in form of jets to rewet hot surface of fuel pin under loss of coolant accident. This kind of design to reflood the fuel bundle is different than bottom and top spray reflooding practiced in PWR and BWR type of nuclear reactors. There are two different kinds of rewetting found during radial jet induced cooling. The first one is due to radial jet impingement and the second one is due to falling film which is below the jet impingement point. Rewetting velocity has been predicted along the length of fuel pin due to radial jet impingement cooling. Temperature of FPS has been varied from 400°C to 700°C with help of different powers supply, simulating decay heat of reactor. A variation of coolant radial jet mass flow rate is from 0.5 lpm to 1.8 lpm. It is considered during ECCS injection. It has been observed from the experiments that rewetting velocity decreases with increasing the clad surface temperature and increases with increasing the coolant mass flow rate. The rewetting velocity in falling film is found to be nearly 1.8 times higher than rewetting velocity predicted in circumferential direction. Further, it is found that maximum heat transfer coefficient increases with increasing the radial jet coolant mass flow rate. The maximum heat transfer coefficient in case of radial jet impingement is found to be nearly 1.5 times the falling film rewetting. Developed correlation predicts the maximum heat transfer coefficient with experimental data well within the error band of ±10%.

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.

Loss of coolant accident (LOCA) along with the unavailability of emergency core coolant system (ECCS) may cause severe damage in the channel like slumping of the fuel pins inside the pressure tube (PT). The thermal behavior of the coolant channel under this condition is a major concern for the safety of the nuclear reactor. To understand this phenomenon, an experimental and numerical study of an Indian pressurized heavy water reactor (PHWR) housing disassembled fuel pins has been performed. The fuel bundle of 37 pins configuration was used. The present investigation is aimed to capture the steady-state temperature profile over a clad tube, pressure tube and calandria tube under completely voided condition. ANSYS Fluent 19.0 was used to carry out the numerical analysis. Close agreement was obtained between experimental and numerical results. Results showed that a significant temperature gradient was developed in the PT and the fuel pins of the outer ring that were in contact with the PT. It was also found that the moderator acts as a good heat sink as it removes 89% decay heat.

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.

A reactor shutdown would typically initiate with a trip signal following a postulated Loss of Coolant Accident (LOCA). The Emergency Core Cooling System (ECCS) assures fuel cooling by flooding the core and recirculating coolant inside the channels for long-term decay heat removal from the fuel. When the primary flow is interrupted, and simultaneously the ECCS is severely impaired or failed, fuel cooling tends to degrade, leading to the heat-up of fuel and coolant channels. Analysis indicates that further progression of such accidents in Pressurized Heavy Water Reactors (PHWRs) is primarily delayed due to the huge water mass in the calandria vessel around the reactor core. However, the horizontal orientation introduces several phenomenological complexities introducing complexity in formulating the Severe Accident Management Guidelines (SAMGs). For a postulated unmitigated low-pressure accident scenario, the accident progresses with the channels heating up, and creep sagging under thermal load and channel and fuel weight. The sagged channel contacts the channel below, transferring heat and its load. With further progress, it is postulated that the supporting channel would fail, and leading to a cascading core collapse. The present paper discusses the experimental studies carried out on multiple channel sagging and disassembly phenomena envisaged during the mid-phase of a severe accident. It is challenging to perform full-length experiments; hence, scaled-down channels were used in the experiments. The postulated accident condition is a significant break LOCA with loss of ECCS and loss of all heat sinks. The experimental setup consists of three channels placed vertically in a fixed lattice pitch forming three distinct channel rows. The paper compares two possible situations or test cases: 1) all three channels are heated, 2) the top two channels are heated, and the lower channel is unheated. The lower channel is an unheated (dummy) channel supporting the upper deforming channels. Finally, a multi-channel sagging deformation mode has been postulated based on the experimental observations. The contact propagation along the contacting surface is found to be a time-dependent event, indicating a channel's rate-dependent elongation. The middle CT is compressed between the upper and lower channels, compromising its structural support for middle PT.

The safe operation of the reactor is compromised under postulated accident events. The events may originate from single or multiple failures. LOCA, along with the unavailability of ECCS, is one such multiple failure event. Phenomena like fuel heat up, fuel slumping, clad oxidation, hydrogen generation and fission product release are about to take place. However, moderator cooling to the external surface of fuel channels limits the fuel heat up. The steam starvation in the channel limits hydrogen generation. The purpose of this study is to analyze the thermo-chemical behavior of a fuel channel of PHWR, assuming that the 37 fuel elements of a bundle are collapsed and collected in a tightly packed stack at the bottom of the pressure tube of the fuel channel. The temperature distribution is obtained under a pseudo-steady-state for the various components of the simulated fuel channel exposed to an oxidative environment. Experimental results have indicated that a significant temperature gradient gets established in radial, circumferential and axial directions in the fuel channel due to slumped fuel bundle configuration as well as hydrogen generation. Estimation from the heat balance studies shows that moderator functions as an active heat sink.

In this paper we have developed a numerical methodology which is capable of estimating volatile fission product release from nuclear fuel under changing irradiation conditions with incorporation of all physical phenomena’s. The present study bridges the gap between previous analytical studies by providing estimation of unstable fission product release under accidental condition with and without incorporation of precursor’s history effect. Based on the present methodology a computer model “FIPRAP-Fission Product Release Analysis Program” has been developed and validated with standard fission product release problems and benchmarked against internationally accepted CORSOR-BOOTH model as well as Accident Source Term Evaluation Code (ASTEC). The result of the proposed model lies within 15% of internationally accepted models.

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.

Moderator acts as primary heat sink to arrest the unwarranted heat-up of fuel channels in the scenario of loss of coolant accident along with the unavailability of emergency core cooling system. In addition, severe heat-up of channel (which consists of fuel bundles, Pressure Tube, PT and Calandria Tube, CT) also occurs during any un-called for event of moderator-cooling/make-up system malfunctioning. Eventually, this all lead to the major loss in structural rigidity. Present investigation is performed to study the sagging deformation of PT for standardized 220MWe Indian PHWR. A scaled-down PT (1/3rd scale) is tested with a uniquely designed experimental setup that is capable of capturing high temperature sagging behavior. Results confirm that the Indian PT significantly deforms under thermally assisted rate-dependent plasticity (creep) above 600 °C. Post-test visual examination reveals no cracks in the PT, however a significant thinning of the PT near the junction of weight simulators is observed.

Establishment of rewetting of hot nuclear fuel pins under a pipe break event is an important design aspect of Nuclear Power Plant (NPP). The successful rewetting of heated pins through Emergency Core Cooling System (ECCS) injection stands to be a major requirement as this action mitigates the accident progression. In Advance Heavy Water Reactor (AHWR), ECCS is designed to inject water from a central water rod of the fuel bundle in form of jets to rewet hot surface of fuel pin. This kind of design to reflood the fuel bundle is different than bottom and top spray reflooding practiced in PWR and BWR type of nuclear reactors. The success of the proposed ECCS injection type has been assessed in a separate effect test where the study has been carried out with a simulated single fuel pin. Influence of issuing radial jet on circumferential and axial conduction and associated rewetting pattern has been studied. As ECCS injection may take place at different decay power levels and at different fuel initial temperatures hence experiments are conducted at different power levels (212 W, 600 W, 780 W) and at different initial clad temperatures (400 °C, 600 °C, 660 °C). The injection flow rate has been kept constant at a design flow rate of 1.8 lpm for all the experiments. In all the experiments it is observed that a steep circumferential temperature gradient is established at the beginning of injection and the gradient comes down with the time advancement. A maximum gradient of 500 °C is observed for a short period of time (∼4 s) for the initial temperature of 660 °C. The influence of variation of jet velocity arising from height of it’s injection is reflected on circumferential rewetting velocities which are found to vary from 2 mm/s to 10 mm/s. A correlation developed from experimental data has been proposed to estimate the average fuel temperature at any time instant during rewetting process. The study concludes that the design of multi point injection within the bundle will successfully be able to rewet the simulated fuel pin over the maximum period of time of ∼28.6 s.

This study aims to assess the applicability of the in-vessel melt retention (IVMR) strategy with external vessel water cooling to the reactors of VVER-1000/v320 type. IVMR strategy is one of the feasible solutions to mitigate reactor vessel failure and further fission products release to the containment and to the environment outside. The reference power plant for this investigation is VVER-1000/v320 reactor sited at Unit 5 and 6 of Kozloduy NPP. The ASTECv2.0r3 severe accident computer code was used to simulate Large Break LOCA (2 × 850 mm) with full Station blackout (SBO) in VVER-1000/v320 reactor model. The external water cooling of the vessel bottom head during the accident was also simulated by the code. The results from four calculations have been assessed: two integral calculations made with ASTECv2.0r3with and without simulation of external vessel water cooling and two stand-alone calculations made with the ICARE module of ASTECv2.0r3. The heat fluxes from the corium to the vessel and the heat fluxes from the vessel to the outside water have been predicted by the code. There were also accounted periods of maximum heat input from the corium to the vessel steel wall.

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.

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.

Under very low probability postulated accident like LOCA with unavailability of ECCS, the moderator heat sink maintains the channel integrity by providing adequate cooling. As the cooling is external to channel, the circumferential as well axial temperature gradient exists within the channel internals. An experimental set-up was fabricated to simulate the boil-off condition during a postulated accident in NPP. The experiments were conducted at average fuel bundle temperature 600 °C, 800 °C and 1100 °C with superheated mixture of steam and argon. The minimum and maximum temperature in CT, PT and fuel bundle were diametrically opposite to each other; the circumferential as well axial temperature gradient increased with an increase in bundle temperature. The rate of moderator heat absorption increases with increase in bundle temperature. The moderator was successfully able to prevent a breach in the channel; it proved to be the ultimate heat sink during a postulated LOCA with unavailability of ECCS.

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.

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.

Loss of coolant accident (LOCA) in nuclear reactor has a low probability to occur, but may have significant consequences due to decaying heat of the nuclear fuel rods. During LOCA with un-mitigated station blackout (SBO) in Indian Pressurized Heavy Water Reactor (IPHWR), there is a mismatch between heat generation and heat removed from the fuel channels, that results in structural failure of the fuel channels. These channels settle down at the bottom of the Calandria vessel, forming debris bed. Decay heat is absorbed by the moderator present in the Calandria vessel. The moderator level decreases due to continuous evaporation, which exposes fuel channels gradually. An experimental setup has been designed and fabricated to simulate the debris bed scenario. Experiments are conducted at different moderator levels, in the Calandria vessel and at different rated power. The temperature of the exposed rods gets stabilized with time due to steam cooling. It is observed that, the temperature is well below the melting point of the cladding, PT and CT material.

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.

Rewetting of a hot surface is the process of establishing direct liquid contact with a large portion of the surface whose initial temperature exceeds that required to maintain film boiling for prescribed surface and flow conditions. The Advanced Heavy water Reactor (AHWR) is a natural circulation vertical pressure tube type boiling light water cooled and heavy water moderated reactor. In case of a loss of coolant accident, the clad surface temperature goes up very high and comes down due to coolant injection from the Emergency Core Coolant System (ECCS). The rewetting takes place in Boiling Water Reactors (BWR) due to top flooding and Pressurised Water Reactors (PWR) due to bottom flooding. But in AHWR, the emergency coolant enters into the reactor core in radial direction and after this, cross flow phenomenon takes place from one fuel pin ring to next. The study is being carried out on the effect of cross flow on rewetting of AHWR fuel bundle. This paper will discuss the modeling of the experimental setup having pressure tube, fuel cluster, steam generator, accumulator etc and study the effect of radial flow on rewetting of fuel pins. An analysis of the model, considering with and without cross flow, has been carried out and shows that the pick fuel temperature is sensitive to cross flow. The thermal hydraulic safety analysis code Relap5/3.2 is being used for modeling of experimental setup for rewetting study.

Experiments have been carried out to study the heatup behavior of a submerged debris bed constituted of segmented reactor channels of Pressurized Heavy Water Reactors (PHWR). The study has been done for under fully submerged, partially submerged and nearly exposed conditions. This situation may arise from a severe accident scenario of PHWR where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure tube and calandria tube constitute each segmented reactor channels. Ten such reactor channels constitute the debris bed. Heatup of this debris bed is observed with respect to different water levels and at power levels of 10, 20 and 30 kW, typical to decay power levels of 0.25 %, 0.5 % and 0.75 % of reactor full power respectively. It has been observed from the set of experiments that fully submerged debris bed does not get heated up and fuel rod temperature remains at saturation temperature. Heatup is observed for exposed channels only, however the heatup rate is limited with steam and radiative cooling. For exposed debris bed, channels surrounded by neighboring channels get heated up as compared to channels at the periphery of the debris bed.

In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2.0 MPa and 4 MPa. In order to simulate partially voided conditions inside PT, asymmetric heating has been carried out by injecting power to selected heater pins of the upper section of the 19 element fuel bundle simulator housed in a PT. This simulates nearly a stratification level of a half filled reactor channel. Through this technique an expected maximum circumferential temperature gradient of around 440 °C, has been attended from top to bottom periphery of PT. Tests also cover a power range of 8–11 kW which simulates different decay power levels. An asymmetric ballooning over eighty percent of PT length is observed for all the experiments and the deformation is mostly located to the upper part of the PT. The PT integrity is observed for lower internal pressure tests however a local failure has been observed for the test at 4.0 MPa. This is found to be due to excessive local strain prior to establishment of contact with Calandria Tube.

Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6–8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment.

In a nuclear reactor, loss of coolant accident (LOCA) considers wide range of postulated damage or rupture of pipe in the heat transport piping system. In case of LOCA with/without failure of emergency core cooling system in a Pressurized Heavy Water Reactor, the pressure tube (PT) temperature can rise significantly due to fuel heat up and gross mismatch of heat generation and heat removal in the affected channel. The extent and nature of deformation is important from reactor safety point of view. Experimental set-ups have been designed and fabricated to simulate ballooning (radial deformation) of PT for 220 MWe IPHWRs. Experiments have been conducted using voided PTs at 4 and 6 MPa internal pressure at different heating rates. It is observed that at 4 MPa and 6 MPa internal pressure, the PT sagged at about 500 °C before the ballooning initiation. The ballooning initiates at a temperature around 625 °C and contact between PT and Calandria Tube (CT) occurs at around 680 °C, respectively, for 4 MPa and the same was at 550 °C and 640 °C for 6 MPa. The structural integrity of PT is retained (no breach) for all the experiments. The PT heat up is found to be arrested after the contact between PT and CT, thus establishing the moderator acting as an efficient heat sink for IPHWRs. A thermal creep model ‘PTCREEP’ has been developed to predict creep behaviour of the PT of IPHWR. It is found that the contact time predicted by PTCREEP is very close to the experimental result. Hence, PTCREEP can be used for the prediction of the ballooning behaviour of the PT for IPHWR in case of LOCA for the operating temperature and pressure range.

Severe accident analysis of a reactor is an important aspect in evaluation of source term. This in turn helps in emergency planning and Severe Accident Management (SAM). The use of the Severe Accident Management Guideline (SAMG) is required for accident situation which is not handled adequately through the use of Emergency Operating Procedures (EOP), thus leading to a partial or a total core melt. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). Initiation of SAMG for VVER-1000 is considered at two core exit temperatures viz. 650 °C as a desirable entry temperature and 980 °C as a backup action. Analyses have been carried out for VVER-1000 (V320) for verification of some of the strategies namely water injection in primary and secondary circuit. These strategies are analysed for a high and low pressure primary circuit transients. Station Black Out (SBO) is one such high pressure transient for which core heat can be removed by natural circulation of the primary circuit inventory by maintaining the secondary side inventory. This strategy has been verified where the feed water injection to secondary side of SG is considered from external power sources (e.g. mobile DG sets) as suggested in SAM guidelines. The second transient, a low pressure event is analysed for verification of the SG flooding and core flooding strategies. The analysis shows that SG flooding is not adequate to arrest the degradation of the core. In case of core flooding strategy, the analyses show that core flooding is not adequate to arrest the degradation of the core for the large break LOCA where as for small break LOCA the injections through available safety systems are adequate. The assessments are carried out with integral severe accident computer code ASTEC V1.3.

The work presented here is an experimental investigation of the critical flashing flow of initially subcooled water through circumferential slits in pipes. The study provides first hand information about the prediction of leak flow rates in piping and pressure vessels retaining high temperature and high pressure. The dedicated experimental facility loop simulates the thermal hydraulic condition of Pressurized Heavy Water Reactors (PHWR). The critical flow characteristics found for varying leakage cross sections at different stagnation pressure and different degree of subcooling has been demonstrated in this paper. A marked decrease in mass flux has been found as subcooling decreases for a fixed stagnation pressure. More observation has revealed that the tighter slits or openings with very short duct as small as 0.8 cm flow length have different flow behavior than greater opening dimensions or with longer flow channels or that for nozzles. The critical flow has been seen to occur at higher pressure differentials along the flaws and prominent changes in the flow rate is reported to occur with varying dimensional parameters of the slit or cracks.

In nuclear reactor probabilistic safety analyses (PSAs), risk is usually defined by the frequency and magnitude of radioactive releases to the environment ( Generic CANDU, 2002 ). An integrated Level-1, -2 and -3 PSA have been carried out for thorium based natural circulation driven advanced heavy water reactor (AHWR). A Level-1 PSA models accident sequences up to the point at which the reactor core either reaches a stable condition or becomes severely damaged, releasing large amounts of radionuclides into the containment. The probabilistic aspects of the analysis focus on the performance and reliability of nuclear plant systems and station staff in response to plant upsets. A Level-2 PSA examines severe reactor accidents through a combination of probabilistic and deterministic approaches, in order to determine the release of radionuclides from containment, including the physical processes that are involved in the loss of structural integrity of the reactor core ( Generic CANDU, 2002 ). A Level-3 PSA goes through the short and long term (radiological) effects on the public ( Fullwood, 2000 ). In this study the risk associated with internal events is only addressed. In the first phase, Level-1 PSA has been carried out to identify postulated initiating events (PIEs) which may lead to severe core damage (SCD) for the reactor. In the second phase, a Level-2 PSA examines two enveloping severe accidents through a combination of probabilistic and deterministic approaches and determines the release of radionuclides from containment. In the third phase, a Level-3 PSA is carried out for the transport of radionuclides through the environment and for the evaluation of public health risk for the two scenarios considered. The salient findings are presented in the paper.

The objective of this study is to develop a system, which assists the operator in identifying an accident quickly using ANNs that diagnoses the accidents based on reactor process parameters, and continuously displays the status of the nuclear reactor. A large database of transient data of reactor process parameters has been generated for reactor core, containment, environmental dispersion and radiological dose to train the ANNs. These data have been generated using various codes e.g., RELAP5—thermal-hydraulics code for the core. The present version of this system is capable of identifying large break LOCA scenarios of 220 MWe Indian PHWRs. The system has been designed to provide the necessary information to the operator to handle emergency situations when the reactor is operating. The diagnostic results obtained from ANNs study are satisfactory.

Pressure tube (zirconium 2.5 wt.% Nb) serves as a pressure boundary for the coolant that removes nuclear heat generated in the reactor core of Indian Pressurised Heavy Water Reactors (IPHWRs). Under postulated low frequency ( −6 per year) accidents like Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS) injection, heatup of pressure tube (PT) combined with internal pressure and the weight of the fuel bundle may lead to deformation. The extent and nature of deformation is important from reactor safety point of view. An experimental set-up has been designed and fabricated to simulate sagging (downward deformation) of PT due to its own weight and the weight of fuel bundles for 220 MWe IPHWRs. Experiments are conducted at different heatup rates of voided PTs. It is observed that sagging initiates at a temperature around 450 °C. Contact between PT and calandria tube (CT) occurs at around 585–625 °C, respectively. Once PT–CT contact takes place, PT temperature either decreases or the temperature rise remains controlled whereas CT temperature keeps on increasing for next 20–30 s. The contact location in all the experiments was near the centre of the tube. Structural integrity of PT is retained (no breach) for all the experiments. The PT temperature rise is found to be arrested after the contact between PT and CT, thus establishing that moderator acts as an efficient heat sink for IPHWRs.

Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.

Severe accident studies for very low frequency events for VVER-1000 (V320) are carried out to estimate in-vessel damage progression under steam-rich and starved conditions. The analyses with code ASTEC, jointly developed by IRSN (France) and GRS, Germany), have shown the influence of steam environment on core heat-up followed by material relocation, hydrogen production, vessel failure and aerosol generation along with release to containment. Hydro-accumulator injection for studied transients also gives rise to a steam-rich environment enhancing the material oxidation depending on the injection time and period. The generated information along with PSA-Level 2 is helpful to decide Plant Damage State (PDS) and fruitfully develop accident management strategies for the plant.

In the case of a loss of coolant accident (LOCA) with/without loss of emergency core cooling system (ECCS), in a PHWR, the pressure tube temperature could rise significantly depending on the scenario due to fuel heat up, with the system internal pressure also varying from 9 to 0.1 MPa during the event. This transient can cause metallurgical and geometrical changes in the pressure tube. Depending on the temperature reached, its structural integrity needs to be ensured during such events with the possibility of pressure tube–Calandria Tube contact. At high temperature the pressure tube deforms plastically due to internal pressure and fuel weight. Both ballooning and sagging behaviour are observed in the pressure tube. However, the relative contribution of each deformation is dictated by the pressure and temperature transient. Pressure tube sagging would always lead to an initial bottom contact. In such cases, studies have shown that the temperature difference between the top and bottom of the pressure tube could be as large as 500 °C. Along with the internal pressure, the thermal stress developed may threaten the integrity of the pressure tube. Modelling has been carried out to simulate the simultaneous sagging and ballooning deformation behaviour expected during the channel heat up condition. This paper describes the prediction of creep deformation in the context of the pressure tube configuration in Indian PHWR's and the study of the simultaneous ballooning and sagging behaviour of a pressure tube. The paper also presents the application of this model for a specific postulated LOCA in Indian PHWRs.

The paper addresses the issue of cold pressurisation during a Non-Loss of Coolant Accident (non-LOCA) and steam line break events. Cold pressurisation is usually avoided during reactor start up from cold and shut down to cold but it has to be avoided even during accidents so that consequential failures do not occur. A non-LOCA event, such as isolation of containment on a spurious signal and two main steam line breaks have been analysed. It is observed that the transient behaviour of the system is similar in these events. This paper discusses the issue of cold pressurisation in these events.

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.