Indian Pressurized Heavy Water Reactors Research Articles

Numerical investigations on the performance of coupled NCL based passive heat removal system of an IPHWR

Passive heat removal systems (PHRS) are indispensable additional safety features in nuclear reactors to deal with station blackout (SBO) conditions. These systems are based on coupled natural circulation loops (C-NCL) connected in series configuration. Further, these system's design, operating conditions, working fluid, heat removal capacity, etc., are specific to the reactor type. In the present work, the operational performance of such first of a kind PHRS of 700 MWe Indian pressurized heavy water reactor (IPHWR) is investigated. For this purpose, a coupled two-phase NCL code is developed, validated and employed. This PHRS consists of two NCLs connected in HHVC-VHVC (horizontal heater and vertical cooler - vertical heater and vertical cooler) configuration. In this study, initially, the full power reactor thermal-hydraulic conditions are established, subsequently, the performance of the PHRS under SBO is investigated. The analyses demonstrated that the system is adequate to effectively eradicate the decay heat by establishing single and two-phase natural circulation (NC) flows of ∼4.5 % and ∼5.6 % of rated conditions in primary and secondary transport loops, respectively. The maximum PHRS capability is predicted to be ∼72 MWth which is more than 3 % of full power. The transients of flow, pressure, energy transfer and heat sink thermal conditions are analyzed when the system changes from forced to NC. Further, the influence of system conditions such as loop diameter, loop height, sink volume, sink temperature and heat exchanger area on PHRS performance was predicted and the changes in system parameters were analyzed. The vital parameters influencing the PHRS performance are identified based on sensitivity analyses. This investigation provides significant insight into the flow and heat transfer characteristics of two-phase and coupled NCLs in the PHRS.

Operating Experience of Snubbers in Indian Pressurized Heavy Water Reactors

In a Nuclear Power Plant (NPP), the piping layout needs to have flexibility to reduce the operating stresses and nozzle loads due to thermal expansion. At the same time, the piping and equipment need to be supported adequately to sustain the dynamic loads viz., seismic loads, sudden opening/closing of valves etc. In Indian Pressurized Heavy Water Reactors (PHWRs), Snubbers are provided on nuclear equipment & piping to serve both purposes i.e. allow flexibility for thermal movement during operation and at the same provide requisite rigidity to take care of dynamic loads. Snubbers are subjected to harsh environment during operation including high temperature, radiation, vibration and corrosive condition, which may adversely affect its performance. Various degradations viz., high drag force, shift in dynamic characteristics (locking characteristic), jamming etc. have been observed. Periodic testing of Snubbers is required to ensure the proper functioning during normal operating condition as well as during dynamic event. A dedicated In-Service Inspection (ISI) program for snubbers exists in Indian PHWRs to ensure healthiness/functionality of Snubbers. Efforts have been made to reduce numbers of Snubbers progressively in PHWR units. Numbers of snubbers have been significantly reduced in latest 700MWe PHWR Design. This paper covers the various degradations observed in snubbers, ISI program for snubbers, design modifications, Snubber testing, assessment for fitness for service, replacement of Snubbers in older units and optimization of snubbers.

Structural Integrity Assessment and Fatigue Evaluation of Feed Water Nozzle of 700MWe PHWR Steam Generator

Steam Generator (SG) is a prime component in Indian Pressurized Heavy Water Reactors (PHWRs). The Steam Generator consists of an inverted vertical U-tube bundle in shell for primary flow (Heavy water). The shell side is designed for recirculation. Steam Generator is a Safety Class-1 component and designed & fabricated as per the requirements of ASME Section III, Subsection-NB.Secondary side of the steam generator consists of a shell, Feed Water Nozzle (FWN), Steam Outlet Nozzle, moisture separator, drier, tubes and shroud. The principal material of construction for secondary side pressure boundary is of low alloy carbon steel. Secondary side of steam generator is subjected various loadings (Dead weight, pressure, pipe reactions & thermal) during normal plant condition as well as transient conditions including severe transient like crash cool down where in the thermal gradients are much higher. To take care of these thermal transient loadings a special thermal sleeve type of design has been incorporated in the FWN.As mentioned above, FWN is subjected to thermal transients throughout its life and susceptible to high thermal gradient which makes it important to evaluate for structural integrity and fatigue failure. This paper covers the approach & results for structural integrity assessment (stress evaluation) and fatigue evaluation (cumulative fatigue damage factor) of Feed Water Nozzle of 700MWe PHWR SG for various Service levels A, B, D & Design. These service conditions comprise of thermal & pressure transients during operating and accidental conditions. Detailed integrity check performed for all conditions by stress categorization at critical locations and qualify primary & secondary stresses against ASME Section III allowable.

An analysis method to assess in-Calandria corium retention time in PHWRs

In Pressurized Heavy Water Reactors (PHWRs), during a postulated severe accident involving the core collapse, the core internals would fall onto the Calandria bottom and form a debris bed. Under such conditions, the core debris bed would be contained by the Calandria, which also facilitates its cooling by the Calandria vault water. In absence of any accident management action, the Calandria temperatures would rise, causing loss of material strength that may also threaten the structural integrity of Calandria. The gross structural failure of Calandria would limit its ability to further retain the core debris and would jeopardize the in-vessel accident management strategies. The knowledge of structural degradation of Calandria with time till its ultimate failure is important for formulating and assessing the severe accident management guidelines. Assessment of Calandria failure under accident conditions involves several important aspects such as complex thermo-mechanical loadings, temperature dependent elastic-plastic and creep modeling, flexibility of different parts and boundary constraints etc. In view of this, a systematic analysis method based on the sequentially coupled thermo- mechanical analysis has been devised and used for the structural assessment of Calandria. The failure assessment has been carried out as per the specified failure criteria for prominent failure modes including large unbounded deformations and plastic instability, large inelastic strains/ ductility exhaustion and creep-stress rupture. The paper covers various considerations taken into account during the analyses along with few case studies on Indian PHWRs.

Fretting wear assessment in PHWR pressure tubes using transient dynamic FEA and Archard’s wear model

In Pressurized Heavy Water Reactors (PHWRs), the fuel bundles are placed in a pressure tube (PT). These bundles rest on inside of pressure tubes through the bearing pads. The pressure tubes are part of the primary pressure boundary of the reactor, and are crucial from functionality and safety point of view. Any phenomena/ factors that may cause structural degradation of PT and may affect its structural integrity, requires thorough investigation. The fretting of pressure tube at bearing pad locations is one such phenomenon, wherein the local material removal due to fretting may lead to stress concentration and associated structural degradations. Therefore, a detailed assessment of the PT fretting damage incurred during its intended life is required. Present paper covers the premise and methodology based on Archard’s wear model for evaluation of the volumetric wear rate of pressure tube. A case study has been done to evaluate the fretting wear in pressure tubes of a large Indian PHWR. 3D transient computational fluid dynamic analysis and finite element analysis with contact modelled between the fuel bundle bearing pads and pressure tube are carried out to evaluate the contact force and sliding distance time histories. Based on the analyses results, the work rate and average saturated fret depth are evaluated. The assessment revealed insignificant concern owing to the fretting of pressure tube for the considered cases.

Analytical model development for investigation on hydrogen management studies in the containment of a nuclear reactor

Containment is the ultimate physical barrier to prevent the spread of active air-borne fission products as well as radiation shielding in normal operation as well as accident condition and is designed to enclose whole reactor systems. The design of containment system for Fleet mode of 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) incorporates a Primary Containment (PC) of pre-stressed concrete enveloped by a Secondary Containment (SC) of reinforced concrete. For the specification of the design pressure of the containment system, analysis is performed for various postulated accidents such as Loss of Coolant Accident (LOCA) and Main Steam Line Break Accident (MLSB). Further Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian Pressurized Heavy Water Reactors (IPHWR) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching local detonation and global deflagration during such a scenario.In order to arrive at the design basis of containment system and simulation of large numbers of accident sequence a computer code is required, where mitigating features such as Passive Auto-Catalytic Recombiners, Containment Spray System & Containment Filtered Venting System can be modeled. A system thermal hydraulic computer code has been developed for containment response calculation during normal operation as well as accident conditions including, severe accident to include the new models for mitigating features. This article presents the details of various models of computer code PACSR-SI-2.0, its validation with data observed for hydrogen distribution & removal in Recombiner Test Facility,Battelle Spray Test for containment system & Spray test at IIT Mumbai.

Critical heat flux model for heated horizontally oriented cylindrical vessel and its application to PHWR calandria under severe accident scenario

In-vessel retention (IVR) of molten corium is a key severe accident management strategy adopted in pressurized heavy water reactors (PHWRs) and is recognised as the only option to manage a complete core melt scenario. Since the PHWR calandria vessel is completely surrounded by a large pool of water, the vessel itself is expected to act as a ‘core catcher’. The success of IVR strategy depends up on whether the actual heat flux imparted by molten corium is less than the critical heat flux (CHF). In the present work, a hydrodynamics model is developed and applied to obtain the variation of CHF along outer surface of 220 and 700 MWe Indian PHWR calandria vessel and tube. The thermal margin available for success of IVR strategy is evaluated. The governing equations for external buoyancy driven boundary layer flow are derived and solved in non-dimensional form. The details of the mathematical model, benchmarking exercises, validation aspects and application of the model to PHWR calandria vessel and tube are discussed. Parametric studies are presented to bring out the influence of diameter, system pressure, subcooling of water, depth of submergence and slip ratio on the variation of CHF along the cylindrical vessel surface.

Development of a coupled neutronics-CFD system for 540 MWE Indian Pressurized heavy water reactors (IPHWRs) and verification with slow online refuelling transient

A loosely coupled reactor is subjected to local perturbations due to reactivity device movement that affects the core neutronics in a local region of the reactor core. It is a challenge to predict the fuel-coolant characteristics like power, enthalpy, temperature etc., from a purely neutronics point of view hence a coupled reactor physics-CFD code is a must for the reactor. Additionally, the continual change in core composition due to on-power refueling along with coolant feedback throughout change in reactor power and/or reactivity device movement makes it clear that development of a coupled code is not sufficient in itself. The coupled code should capture the burnup effect of the fuel and the change in fuel heat transfer characteristics due to fission gas accumulation in the plenum gap. In this work a cross-section library with small step cross-section data for fuel and coolant has been developed. Simple interpolation gives a reliable data for neutronics computations. The coupled code discussed in this paper is an attempt to capture fuel characteristics accurately using a small step cross-section data and explicit computation of the effect of plenum gas in the fuel. The neutronics calculation is carried out in modified TRIVENI code and the CFD computation is carried out in Openfoam. Recent trends in reactor physics is towards high fidelity computing using Monte Carlo for pin-resolved neutron transport coupled with thermalhydraulics code. However, these high fidelity codes are computer resource intensive where as our code can be run on meagre computing resources. There is scope for improvement in not so well understood phenomena like plenum gas behaviour on the fuel characteristics and this work is highlighting this aspect of PHWR fuel.

Numerical Investigation on Heat Flux Variation in Fuel Bundle Over Fully Voided Concentric Channel of IPHWR Under Severe Accidental Scenario

Abstract A steady-state thermal analysis is performed on a fully voided channel of an Indian pressurized heavy water reactor (IPHWR) by providing variable heat flux inputs of 800 W/m2, 1500 W/m2, 2500 W/m2, 3500 W/m2, and 4500 W/m2 for evaluating the numerical results of temperature distributions in pressure tube (PT) and calandria tube (CT) at concentric scenarios using a numerical approach. The top to bottom temperature difference of PT and CT is found to be 0.57% and 3.31%, respectively, for 800 W/m2 input heat flux, and this variation decreases to 0.015% and 0.05%, respectively, for 4500 W/m2 input heat flux, thus pointing out that increasing the heat flux leads to the decreased top to bottom temperature difference. Also, the temperature distribution pattern is found to be similar for all the input heat fluxes, and for a particular heat flux, the change in circumferential temperature is found to be negligible. The boiling phenomenon starts at 2500 W/m2, and the volume fraction of moderator vapor increases as the heat flux is increased, the average density of moderators inside the enclosure decreases, and continuous variation in film thickness is also observed with respect to time for a particular heat flux value. The results also showed that the moderator acted as an effective heat sink for higher heat input rates.

Experimental assessment of hydrogen removal rate for passive auto catalytic recombiners

Generation and accumulation of hydrogen in containment atmosphere could pose a potential threat to the integrity of containment as hydrogen can form flammable mixture with air in the containment during postulated severe accident scenario involving multiple failures. Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian pressurized heavy water reactors (PHWRs) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching deflagration/ detonation limits during such a scenario.Out of several options available for hydrogen mitigation, Passive Auto-catalytic Recombiner (PAR) is one of the most effective and recommended (IAEA -TECDOC-1661(2011)) technology for the removal of hydrogen inside the containment. The testing and evaluation of the performance of PAR for different conditions envisaged in the containment (such as pressure, temperature, steam concentration and poison) is a vital requirement for the deployment of these devices in the nuclear power plants (NPPs). The behavior of catalytic recombiners in terms of hydrogen removal rate with varying hydrogen concentrations is required to be evaluated essentially to work out the numbers and location of recombiners inside the containment. Hence, for testing of catalytic recombiners, an instrumented test facility simulating different reactor containment environment under controlled conditions, has been commissioned. Recombiner Test Facility (RTF) is a test rig for testing of full scale prototypes of catalytic recombiners.Results of Hydrogen Removal Rates (HRR) for varying hydrogen concentration, method for the calculation of HRR and its validation, model equation developed for Corrected Hydrogen Removal Rates (CHRR) usable in numerical simulations are presented in this article. With the proposed method for CHRR, it was established that the Recombiner performance can be evaluated.

Numerical analysis of thermal-hydraulic safety for a reactor containment in the event of a main steam line break

The present work demonstrates the development of a numerical thermal-hydraulic (TH) model for the containment module of the PRABHAVINI code which is an in-house severe accident analysis computer code used for the analysis of Indian pressurized heavy water reactors (PHWRs). The model, for the analysis of the TH behavior of the containment under a postulated main steam line break (MSLB) situation inside it, incorporates the effect of nonlinear coupling among different heat and mass transfer chronologies such as heat transfer within the containment and through its walls, wall and interface condensations, condensation due to over saturation of the air and evaporation.The TH model accounts for the thermodynamic non-equilibrium between the air–vapor mixture inside the containment and the water likely to be accumulated at the bottom of the containment.For the comparison of the TH model developed, a more conservative case study is undertaken where a single containment is analyzed for an MSLB accident ceasing the possibility of any kind of leakage from it. The results obtained from the model developed are compared with those obtained from the RELAP5. Next, the model is extended for more realistic TH analysis of the multi-volume containment incorporating the possibility of uni- or bi-directional inter-volume leakages as well as leakages from the containment volumes to the atmosphere. The analysis demonstrates the effect of different heat and mass transfer phenomena on the critical TH field variables of the reactor containment for the MSLB situation under consideration.

Anisotropy and variability in thermal creep behaviour of Zr-2.5Nb pressure tube

The Zr-2.5Nb alloy pressure tubes of Indian Pressurized Heavy Water Reactors were manufactured from double melted ingots initially and subsequently from quadruple melted ingots to reduce impurities such as chlorine, phosphorus and carbon. In the current investigation, creep tests were carried out in the stress range of 0.7–0.9 times the yield strength and the temperature range of 350 °C–450 °C to characterize the thermal creep behaviour of Zr-2.5Nb alloy pressure tubes fabricated from double and quadruple melted ingots. The rupture time, minimum creep rate, stress exponent, activation energy and threshold stresses were obtained by carrying out the creep tests of the samples fabricated with their axis parallel to the axial and transverse directions of the pressure tubes. Experimental analysis revealed that the rupture time for double melted tube was 2–3 times lower than that of quadruple melted tubes, whereas no significant change in minimum creep rate was observed. Stress exponent values were obtained in the range of 3–5 signifying dislocation creep as the dominant creep mechanism. To predict the long-term creep properties, the master curves employing the Larson-Miller and Monkman-Grant methods were also evaluated. The effect of microstructural anisotropy, alloying elements and impurities on the thermal creep behaviour of Zr-2.5Nb pressure tube has been investigated.

Numerical investigations on boiling of moderator during PT/CT contact in 220 MWe IPHWR under postulated accidental scenario

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.

Effect of manufacturing route on thermal creep behaviour of Zr-2.5Nb pressure tube alloy used in Indian PHWR

Indian Pressurized Heavy Water Reactors (IPHWRs) uses Zr-2.5Nb alloy pressure tube (PT) in cold work and stress relieved condition. These PTs are subjected to extreme conditions of stress, temperature and irradiation. Since past three decades of PHWR operation in India, fabrication routes of PTs for IPHWR has undergone several modifications to enhance its in-reactor performance. The deformation of PT in PHWR occurs due to irradiation creep, irradiation growth and thermal creep. In the present study, the different generations of PTs used in IPHWR are tested for its thermal creep behaviour. The accelerated creep rupture tests were carried out using flat tensile specimens with its axis parallel to the longitudinal direction of the PT in the temperature range of 350 to 450°C under stresses in the range of 0.7 to 0.9 times of yield strength at the respective temperatures. The minimum creep rates were evaluated along with the stress exponents and activation energies. The values of stress exponent and activation energy revealed the underlying creep mechanisms responsible for the creep deformation. Microstructural examination was carried out to understand the need and efficacy of the new fabrication route. The observed thermal creep behaviour of all the generations of pressure tubes were rationalised in terms of their respective texture and microstructures.

Experimental investigation of CsI aerosol removal in containment spray system test facility of 700 MWe IPHWR

Post-accident management of the radionuclide in a nuclear reactor is always a focus point from the nuclear power plant safety point of view. A number of systems using different radionuclide scrubbing mechanisms are already deployed in the Indian nuclear reactors to safeguard the public during and post-accident scenario. However, it is always desirable to achieve “As Low As Reasonably Achievable” (ALARA) philosophy. In the same lines, to further strengthen the safety of Indian nuclear reactors, Containment Spray System (CSS) is introduced in 700 MWe Indian pressurized heavy water reactors to cater/mitigate the aftermath of the Design Basis Accidents (DBA) i.e., Loss Of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). It provides the fast vapour suppression and radionuclide removal from containment during/post design basis accidents. In the present work the experimental investigation of the effect of water spray on the CsI aerosol scrubbing/CsI aerosol removal is studied in a scaled down facility. Experiments are also carried out for the CsI aerosol removal when spray is not operating. In order to encompass the containment pressures during DBA, 1.0 bar(g) initial vessel pressure is used to simulate the accident conditions. For simulation of aerosols after accident CsI nano particles (spherical) suspended in the air and CsI dry aerosols are used. The spray flow rate for all the experiment is kept constant approximately 11.3 lpm. However different Sauter Mean Diameter (SMD) (D32 = 590 µm, 745 µm, 1000 µm) are used to study the effect of droplet diameter on CsI aerosol removal. The temporal variation of the vessel relative pressure gas mean temperature and wall temperature is measured. Vessel depressurization rate with time is recorded. CsI aerosol removal/ CsI aerosol concentration variation with time is measured using two diverse measurements system namely scrubber method and Optical Particle Counter (OPC). Global decontamination factor and global collection efficiency is deduced based on the studies. Few tests without spray are also carried out to estimate the plate out effect on CsI aerosol removal.

Radiative heat transfer analysis of 37-pin fuel bundle of Indian pressurized heavy water reactor under heat-up condition: experimental, numerical and analytical study

ABSTRACT 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.

Experimental and numerical study of 19-pin disassembled fuel channel under severe accident condition

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.

An innovative scheme to conserve natural Uranium in Indian Pressurized Heavy Water Reactors

Natural Uranium being a limited resource on earth needs to be conserved and utilized optimally when used in the reactor core. Pressurized heavy water reactors (PHWRs) are fueled with natural Uranium (NU) in the form of short length bundles. Features like online refueling and short length bundles in PHWRs, provide the flexibility in towards optimizing the fuel requirement and the discharge burn-up during any stage of operation. In equilibrium core conditions, nearly fifty percent of fuel bundles have achieved burnup <3.5 GWD/TeU which is half of their target value of ~7GWD/TeU. There are instances like En-masse coolant channel replacement (EMCCR) or planned de-commissioning, where the full core has to be emptied. A small modification in fuel management scheme months before long shutdown activity can help in utilizing the low burn-up bundles present in the core for extended operation of reactor and save NU bundles. In this paper, a novel scheme for refueling and re-shuffling of fuel bundles has been developed based on heuristic approach which is further supported by numerical simulation. This modified in-core fuel management (ICFM) scheme leads to a net economic benefit in terms of saving of fresh NU bundles as the feed rate of fresh NU bundles will be reduced to ~35% of feed rate during normal ICFM.

DESIGN AND INVESTIGATION OF WEIGHT BUNDLE SIMULATOR FOR INDIAN PHWR USING APDL - A THERMAL ASPECT

Under a postulated scenario of Loss of Coolant Accident (LOCA) with un-availability of emergency core cooling system (ECCS) for Indian Pressurized Heavy Water Reactor (IPHWR), the channel integrity needs to be assured. An experimental facility is presently being developed at Indian Institute of Technology Roorkee (IIT R) India to study such a severe event. In this study a CFD simulation of fuel bundle weight simulator is being carried out using ANSYS 19.0 in transient thermal analysis using ANSYS Parametric Design Language (APDL) solver. The test facility aims to estimate the thermal aspect of weight bundle simulator which is the source of heat generation for pressure tube. Thermo-mechanical deformation of pressure tube will depend on the heat source, therefore it is of great importance to check the thermal integrity of weight bundle fuel simulator. Power requirement, weight, and dimensions of weight bundle fuel simulator is similar to the actual fuel bundle used in 700 MWe IPHWR. Simulation was done for 3 % and 2% decay heat. From analysis it was found that the rate of temperature rise in pressure tube with 3 % decay heat reached a maximum of 2 0C/sec and average rate of temperature rise was around 1 0C/sec whereas with 2% decay heat, pressure tube maximum temperature rise was 1.5 0C/sec and average rate 0.8 0C/sec. Results obtained will be further used for designing of 700 MWe full length channel set-up.

Critical Heat Flux on Curved Calandria Vessel of Indian PHWRs During Severe Accident Condition

Abstract In pressurized heavy water reactors (PHWRs), during an unmitigated severe accident, the absence of adequate cooling arising from multiple failures of the cooling system leads to the collapse of pressure tubes and calandria tubes, which may ultimately relocate to the lower portion of the calandria vessel (CV) forming a debris bed. Due to the continuous generation of decay heat in the debris, it will melt and form a molten pool at the bottom of the CV. The CV is surrounded by calandria vault water, which acts as a heat sink at this scenario. In-vessel corium retention (IVR) through the external reactor vessel cooling (ERVC) is conceived as an effective method for maintaining the integrity of a calandria vessel during a severe accident in a nuclear power plant. Under the IVR conditions, it is necessary to ensure that the imparted heat flux due to melt is less than the critical heat flux (CHF) at the bottom of the calandria vessel wall. To evaluate the thermal margin for IVR, experiments are performed in a prototypic curved section of calandria vessel (25o sector) of calandria vessel to determine the CHF, heat transfer coefficient, and its variation along with the curvature of calandria vessel. The effect of moderator drainpipe on CHF and the heat transfer coefficient has also been evaluated. It has been observed that the imparted heat flux is much less than the CHF at the bottom of the calandria vessel.