A new reactor system under design study is a vertical pressure tube-type boiling light water cooled and heavy water moderated. Main heat transport system consists of reactor core, core inlet and outlet extensions, inlet feeders, tailpipes, steam drums, downcomer, and inlet header. It incorporates several advanced passive safety features, e.g., heat removal through natural circulation both during normal operation and reactor shutdown without the use of any coolant pump. Safety remains the paramount consideration in the design, while plant availability is the next. Several reactor trip set points are provided in design to tackle the disturbances in operation, and many safety systems are provided to mitigate different accidental conditions. A major objective of the reactor design is to provide a capability to withstand a wide range of postulated events (PIEs) without exceeding the acceptance limits, thereby maintaining fuel integrity. Partial loss of feed water is one of such PIEs. Following article highlights the modeling and overall transient behavior and its safety implication.

The strong earthquake followed by tsunami, a low-frequency and high-consequence natural disaster, at Fukushima, has shaken the world’s faith in nuclear energy. The new reactor design around the world has started looking ways to mitigate this scenario with innovative design features, whereas every nuclear country has started reviewing their existing reactor response to this scenario minimizing the consequences or limiting the damage progression. India has designed an advanced reactor concept, i.e., advanced heavy-water reactor (AHWR). This reactor has many passive design features to mitigate the consequences not only for the postulated design basis events but also for beyond design basis events. During station blackout (SBO) scenario leading to hot shutdown, the isolation condensers (ICs) are intended to remove decay heat with the help of overhead large pool of water, i.e., gravity-driven water pool (GDWP). This paper outlines the assessment of station blackout scenario for AHWR using the last version of the French best estimate computer code CATHARE2/V2.5_2 and its comparison with RELAP5/Mod3.2 findings. First, it explains the modeling of main heat transport system of AHWR and isolation condenser loop along with GDWP in CATHARE2 followed by thermal-hydraulic safety assessment of station blackout scenario and comparison of predictions with RELAP5 findings.

The maximum surface heat flux is determined during jet impingement quenching of hot vertical stainless steel surface of 800 ± 10 °C initial temperature. Water jet through a sharp edge nozzle of 24 ± 1 °C temperature is injected on a vertical test surface of 0.25 mm thickness. The investigation is made up for stagnation point to 24 mm downstream location above and below of the stagnation point. Water flow is regulated to maintain the jet Reynolds number in the range of 5000–24,000. It has been observed that maximum surface heat flux is highest for the stagnation point and reduces monotonically for the downstream spatial locations. The decrease in maximum surface heat flux is higher for the locations above the stagnation point as compared to the locations below the stagnation point. However, for the entire measured spatial locations the maximum surface heat flux increases with the rise in jet diameter and coolant flow rate. The proposed correlation for the maximum surface heat flux predicts experimental data within an error band of ±15%.

A stainless steel (SS-316) vertical rod of 12 mm diameter at 800 ± 10 °C initial temperature was cooled by normal impinging round water jet. The surface rewetting phenomenon was investigated for a range of jet diameter 2.5–4.8 mm and jet Reynolds number 5000–24,000 using a straight tube type nozzle. The investigation were made from the stagnation point to maximum 40 mm downstream locations, simultaneously for both upside and downside directions. The cooling performance of the vertical rod was evaluated on the basis of rewetting parameters i.e. rewetting temperature, wetting delay, rewetting velocity and the maximum surface heat flux. Two separate Correlations have been proposed for the dimensionless rewetting velocity in terms of rewetting number and the maximum surface heat flux that predicts the experimental data within an error band of ±20 and ±15 % respectively.

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 x 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. (C) 2016 Elsevier B.V. All rights reserved.

A vertical hot rod of 12 mm diameter at 800 ± 10 °C initial temperature has been quenched by sub-cooled round water jet. The water jet of 2.5 and 3.5 mm diameter, at jet Reynolds number of 5000–24,000, impinges normal to the test section of SS-316. An infrared camera is used to determine the wetting front velocity on the hot test surface. The investigations are made up to 40 mm downstream locations in both upper and down sides of the stagnation point. It has been observed that during transient cooling, the wetting front velocity increases with the rise in jet Reynolds number and jet diameter. However, rewetting velocity reduces drastically for the extreme downstream locations away from the stagnation point. The reduction in the wetting front progression for the upper side downstream locations is higher as compared to corresponding bottom side locations. The correlation proposed for the dimensionless rewetting velocity predicts the experimental data of upper and bottom side downstream locations in the error band of +30 to −20 %.

Source Term determination under normal operating and accident conditions is one of the important safety aspect of nuclear reactors as “Source Term” strength directly influence the plant and public risk. The general information is used for licensing the reactor and accident management purposes. The source term of a nuclear reactor depends on several design aspects of the reactor. The fuel composition and enrichment, burnup and specific power decide the initial core inventory of fission products and actinides. The release of fission products from the fuel to the coolant is governed by the accident type where fuel temperature plays an important role. Further several mechanism of retention of fission products in coolant system piping and containment surface and leakage paths governs the release to environment. Study has been carried out to assess the fission products and their activities release to atmosphere during different accidental conditions for Advanced Heavy Water Reactor (AHWR), a thorium fuelled and natural circulation driven reactor. As Source Term estimation depends on the reactor type, its layout, the study encounters several uncertainties in the process of fission product release and transportation post accident. Various uncertainties in Source Term estimation for a reactor depend upon several factors like: 1. Uncertainties in equilibrium incore inventory arises due to calculation based upon fuel average enrichment, average burnup and average exposure time instead of three rings actual fuel composition and burnup and interaction cross-section library for AHWR specific fuel; 2. Uncertainties in radionuclide release from fuel arises due to non-availability of suitable diffusion coefficient model, and lack of fuel grain properties like grain sizes, shape, and grain boundary sweeping behaviour information; 3. Uncertainties in radionuclide transportation (retention and release) within reactor main heat transport (MHT) system arises due to various reasons like: lack of released radionuclides nature (aerosol, or molecular) information and their chemical composition; lack of information about generated radionuclide aerosol properties like shape, size distribution, density and actual injection rate ; due to lack of thermal hydraulic parameters like pressure, temperature, flow rate and environmental conditions; simplification in simulation of complex geometry like fuel bundle, end fitting and steam drum regions. 4. Uncertainties in radionuclide transport within reactor containment mainly arises due to uncertainties in the radionuclide chemical compositions, chemical form of iodine (aerosol to elemental ratio), steam and reducing atmosphere with hydrogen, plate out and partitioning factors as well as on the leakage rates from containment under design pressure and beyond design pressure.

The rapid quenching of a hot surface is desirable in several industrial applications, e.g., metal processing, nuclear power plants, electronics, etc. Therefore, an experimental investigation has been carried out on a hot vertical stainless steel surface of 0.25 mm thickness at 800 +/- 10 degrees C initial temperature. The surface has been quenched with the impingement of a round water jet in the range of 2.5-4.8 mm diameters. The maximum surface heat flux during quenching has been determined for jet Reynolds number in the range of Re = 5000-24,000. The observations are made from the stagnation point to the 24 mm downstream spatial locations, for both upside and downside directions of the test surface. It has been observed that the maximum surface heat flux increases with the rise in jet Reynolds number and jet diameter. The correlation proposed to determine the maximum surface heat flux predicts the experimental data within an error band of +/- 20%. The published correlation for the horizontal surface predicts the experimental data of maximum surface heat flux within the range of +40% to -20%.

been carried out for VVER-1000 V320 reactor following LOCA along with Station-Black-Out using the severe accident code ASTEC. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for three break sizes. Since these predictions are dependent on different core degradation parameters and models, a sensitivity analysis is also carried out to study the effect of different core degradation parameters and models on severe core damage progression.LOCA analyses show a variation of degradation parameters which is a consequence of steam-rich and steam-starved conditions. All these calculations of VVER-1000 scenarios show that the ASTEC V2 code gives reliable and consistent results on this type of reactor, including its specifics like Zr1%Nb clad material and horizontal steam generators. (C) 2013 Elsevier B.V. All rights reserved.

Nuclear power plant experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process support systems etc. In such a situation, the plant may result in an abnormal state which is undesired. In case of such an undesired plant condition, the operator has to carry out diagnostic and corrective actions. When an event occurs starting from the steady state operation, instruments’ readings develop a time dependent pattern and these patterns are unique with respect to the type of the particular event. Therefore, by properly selecting the plant process parameters, the transients can be distinguished. In this connection, a computer based tool known as Diagnostic and Prognostic System has been developed for identification of large pipe break scenarios in 220 MWe Pressurised Heavy Water Reactors (PHWRs) and for prediction of expected “Source Term” and consequence for a situation where Emergency Core Cooling System (ECCS) is not available or partially available. Diagnostic and Prognostic System is essentially a transient identification and expected source term forecasting system. The system is based on Artificial Neural Networks (ANNs) that continuously monitors the plant conditions and identifies a Loss Of Coolant Accident (LOCA) scenario quickly based on the reactor process parameter values. The system further identifies the availability of injection of ECCS and in case non-availability of ECCS, it can forecast expected “Source Term”. The system is a support to plant operators as well as for emergency preparedness. The ANN is trained with a process parameter database pertaining to accident conditions and tested against blind exercises. In order to see the feasibility of implementing in the plant for real-time diagnosis, this system has been set up on a high speed computing facility and has been demonstrated successfully for LOCA scenarios.

Under limited core damage accidents (LCDAs) of Pressurized Heavy Water Reactor (PHWR), coolable geometry of the channel might be retained thanks to the presence of moderator heat sink. Indeed, the pressure tube is amenable to creep deformation at high temperature due to internal pressure and fuel bundles weight. Partial or complete circumferential contact between pressure tube and calandria tube aids heat dissipation to the moderator. A new module has been developed by Bhabha Atomic Research Centre (BARC) for simulating this phenomenon which is specific to horizontal-type of reactors. It requires additional calculation of pressure tube sagging/ballooning and temperature field in the circumferential direction. The module is well validated with available experimental results concerning pressure tube deformation and the associated heat transfer in the area of contact. It is then used in analysing typical LCDAs scenarios in Indian PHWR under low and medium internal pressure conditions. This module is implemented in the ASTEC IRSN-GRS severe accident code version under development and will thus be available in the next major version V2.1.

The rewetting phenomenon was experimentally investigated for the hot horizontal stainless steel surface of 0.25 mm thickness and 800 +/- 10 degrees C initial surface temperature. The round water jet of 2.5 mm diameter at 22 +/- 1 degrees C temperature was injected through a sharp edge nozzle. The investigation was done for the stagnation point to 12 mm (approximate to 5d) radial distance and jet Reynolds number varied in the range of 5000-24,000. The rewetting phenomena during the transient cooling was accessed on the basis of rewetting temperature, wetting delay, wetting speed and maximum surface heat flux.It has been observed that with rise in jet Reynolds number, the rewetting performance increased for the entire measured spatial locations. However, for downstream region the surface rewetting has been delayed and occurred at the reduced surface temperature as compared to the stagnation region. The maximum surface heat flux is the highest at the stagnation point and in a range of 2.1-2.45 MW/m(2). The correlation developed for the wetting front speed predicts 85 percent of experimental data within an error band of +/- 20 percent and the correlation for maximum surface heat flux predicts 90 percent of experimental data within an error band of +/- 10 percent. (C) 2013 Published by Elsevier Masson SAS.

An experimental investigation has been carried out to study the cooling of a hot horizontal stainless steel surface of 0.25 mm thickness, which has 800 ± 10 °C initial temperature. A round water jet of 22 ± 1 °C temperature was injected over the hot surface through a straight tubes type nozzle of 2.5 mm diameter and 250 mm length. The experiments were performed for the jet exit to target surface spacing in a range of 4–16 times of jet diameter and jet Reynolds number in a range of 5000–24,000. The rewetting velocity during transient cooling of hot surface was determined with the help of time variant surface temperature data and with the captured thermal images of the hot surface as well. The effect of Reynolds number, Re, jet exit to surface spacing, z/d, on the rewetting velocity has been determined for the different downstream spatial locations. A correlation has also been developed to determine the rewetting velocity, which predicts 75% of experimental data within an error band of ±10%.

An investigation has been carried out to investigate the effect of nozzle geometry on hot horizontal surface rewetting during water jet impingement cooling. The test surface of 800 ± 10°C initial surface temperature is cooled by water jet of 22 ± 1°C temperature. The water flow is varied to maintain the jet Reynolds number in a range of 5,000 to 24,000. The rewetting phenomena with sharp-edged and tube-type nozzles are compared on the basis of rewetting temperature, wetting delay, rewetting velocity, and maximum surface heat flux. The rewetting performance with tube-type nozzle is better than the sharp-edged nozzle particularly for the downstream spatial locations; however, maximum surface heat flux at the stagnation region is higher with the sharp-edged nozzle.

In pressurized heavy water reactor (PHWR), under postulated scenario of small break Loss of Coolant Accident (LOCA) coincident with the failure of Emergency Core Cooling System (ECCS), a situation may arise under which reduction in mass flow rate of coolant through individual reactor channel can lead to stratified flow. Such stratified flow condition creates partial uncover of fuel bundle, which creates a circumferential temperature gradient over PT. The present investigation has been carried out to study thermo-mechanical behaviour of PT under asymmetric heating conditions fora 220 MWe PHWR. A 19-pin fuel simulator has been developed in which preferential heating of elements could be done by supplying power to the selected pins. The asymmetric heating of PT has been carried out at pressure 2 MPa and 1 MPa, respectively, by supplying power to upper region heating elements thus creating an half filled stratified flow conditions. The temperature difference up to 425 degrees C has been observed along top to bottom periphery of PT. A comparison is made between thermo-mechanical behaviour of PT under asymmetrical and symmetrical heat-up, expected from a large break LOCA condition. The radial expansion rate during symmetrical heating is found to be much faster as compared to that for asymmetric ballooning of PT at the same internal pressure. Integrity of PT is found to be maintained under both loading conditions. Heat sink around of test section, simulating moderator is found to be helpful in arresting the rise in temperature for both fuel pins and PT, thus establishing moderator as an effective heat sink under accident conditions. (c) 2012 Elsevier B.V. All rights reserved.

A stainless steel surface of 0.25 mm thickness at 800 +/- 10 degrees C initial temperature was quenched by the jet impingement cooling method. The transient surface heat flux was determined during surface quenching with three different jet diameter The flow of water at 22 +/- 1 degrees C temperature was regulated to maintain the jet Reynolds number in a range of 5000-24000. The observations were made form the stagnation point to the 12 mm downstream spatial location. The quenching performance of the horizontal test surface was evaluated on the basis of maximum surface heat flux obtained during transient cooling.It was observed that the maximum surface heat flux increases with the rise in jet diameter and became the highest for the largest jet diameter at Re = 24000. The correlations developed to determine the maximum surface heat flux was able to predict the experimental data well within the error band of +/- 15%. (C) 2013 Elsevier B.V. All rights reserved.

The SARNET activities on Molten Corium Concrete Interaction aim finally to bring the research results to real reactor applications. These activities concern in particular consolidation of the applications of the ASTEC integral code to the simulation of ex-vessel physicochemical processes during MCCI at real NPPs, such as a VVER 1000 reactor type. This paper presents some of the main issues that appear during the MCCI analyses on VVER1000 test cases. The main observation is that due to the high iron content in the concrete pool stratification with metal below appears early promoting axial ablation and continues to exist till the basemat melt-through time even in case with water reflooding. Axial concrete basemat melt-through appears early due to early melt pool stratification with metal layer below the oxide layer. Higher thermal heat conductivity of the metal layer results in faster concrete ablation. The early melt-through time can be explained also by the impact of high oxide/metal convective heat transfer. It could be mentioned that the efficiency of top water reflooding is influenced by permeability of the pool upper crust. Sensitivity parametric investigations and additional analyses on VVER 1000 MCCI benchmark are presented in this paper. One of them considers the lateral metallic door melting and the transfer of corium to the neighbour cavities of the containment. Additional MCCI studies by UJV on VVER 1000 (type 320 with different containment – concrete door in the tunnel) with various codes as ASTEC (MEDICIS module), MELCOR (CORCON module) and CORQUENCH made are also included hereafter to emphasize some important issues of MCCI phenomenology. They are focused on the assessment of the effect of melt pool stratification on the progression of the ablation front and on the effect of top water cooling.