• R&D pertaining to structural, thermal-hydraulics and probabilistic safety of Indian NPPs
• Fracture, fatigue & creep assessment of nuclear structural components
• Development of severe accident management guidelines and computer code PRABHAVINI
• Thermal-hydraulic safety analysis for Indian NPPs
• Multi-scale material modeling to determine mechanical properties of aged materials
• Core and containment safety studies
• Probabilistic safety analysis and reliability studies for nuclear facilities
• Seismic and civil design of structures of nuclear facilities
The objective of this work is to design of suitable test setup to load alloy 690 specimens in high temperature liquid glass environment and study the change in mechanical properties. Tests on alloy690 specimens have been conducted for 42 to 120 hours of duration at different stress levels (20 to 50 MPa) and the test temperature is 800 to 900 deg. C. When compared to unexposed data, the mechanical properties degrade with time of exposure to corrosive liquid glass.
Due to small wall thickness of the pressure tubes, designing and machining of specimens in all the three directions, especially in the radial direction, is a challenge. By following a novel design, various specimens were tested to study the anisotropy in plastic deformation and hardening properties. Similarly, the creep deformation behavior is also studied by conducting experiments from 500 to 1000 deg.C.
It is known that the state of deformation as well as stress under the ball is multi-axial in nature. In this work, a new procedure for estimation of equivalent stress and plastic strain during the process of indentation in the most stressed location beneath the ball has been developed. This method can be applied to determine the material stress-strain curve of aged materials.
In reality, the power plant components are subjected to multiaxial non-proportional loading. The multiaxial loadings significantly affects the fatigue life and these are not adequately accounted in the fatigue models current design. By using large number of experiments and in-house multiaxial fatigue and cyclic plasticity investigations, new models for improved fatigue design / life and cyclic plasticity assessment were developed and validated.
A computer program for Probabilistic Structural Integrity Analysis of Reactor Pressure Vessel (BARC-PROSIAR) was developed. This program is useful for life extension of the Reactor Pressure Vessel of a light water reactor. Code estimates pressure stress and thermal stresses for the given thermal transient by accounting various uncertainties such as fracture toughness, flaw size etc.
In AHWR, Small Break LOCA scenario poses a challenge as low pressure injection does not actuate since system does not depressurize enough. Passive Auto Depressurization System (PADS) has been designed to establish flow path between IC and Steam Drum (SD) based on low SD level. High pressure signal is generated in PADS to actuate the passive valve. Feasibility and Validation of Passive Pressure Pulse Generator Experimental set-up for Auto Depressurization has been carried out.
A simplified beam macro-model is developed by using modified hinge properties taking into account the reduction in bond strength, loss of rebar diameter and reduction in mechanical properties of steel due to corrosion effect and is validated with beam bending tests. The proposed beam model is used for capacity evaluation of 21year old corroded RC structure and 60year old corroded RC jetty structure. Finally, the service life is predicted for the RC structures from capacity evaluation and NDT tests conducted on the structures.
The objective of the present work is to develop an improved design of field shaper that may withstand multiple use, thus, resulting in a cost reduction of the EMPW process. In this work, advanced computational analyses are carried out to simulate the magnetic fields and resulting deformations in a field shaper. Structural deformations in the existing design of field shaper are computed and numerical predictions are in good agreement with the experimental results.
Sequentially coupled thermal- hydraulic and structural analyses were conducted to assess the structural response of Calandria under a postulated severe accident. Calandria failure time due to plastic instability, excessive inelastic strains and creep-stress rupture was evaluated for standard 220 MWe and 540 MWe PHWR. High temperature tensile and creep properties of Calandria material were generated for temperatures up to 1100 deg.C
Design basis reports of seven buildings of AHWR viz. Reactor Building, Fuel Building, Control Building, Service Building, Station Auxiliary Building, Backup Control Building and Diesel Generator Building considering all the design loads are prepared. FE models are generated and static along with seismic analysis is carried out. Floor Response Spectra are being generated on each floor of all the seven buildings by performing Time history analysis.
Shake table testing of RC structures is carried out for incremental increasing earthquake excitation till failure of the structure is achieved. The results of the tests are utilized to study the failure pattern of RC structures subjected to high earthquake excitations and to develop Nonlinear Floor Response Spectra. Moreover, shake table tests on glove boxes are carried out in order to assess its stability, the structural integrity and leak tightness when subjected to earthquake loads.
Design Basis Ground Motion (DBGM) parameters are generated for BARC-Vizag and AHWR-Tarapur sites using Probabilistic Seismic Hazard Analysis with Multi Expert Participation (PSHA-MEP). Uniform Hazard Response Spectrum (UHRS) are genearted for different return periods of the site.
In the Structural and Seismic Engineering Section (SSES), Seismic qualification of new facility and Re-qualification of old facility for all safety related system, Piping, Mechanical components are carried out for design loads such as dead weight, thermal load, seismic, flow induced vibration (FIV) and wind loads. Shake table testing is also performed for Seismic Margin Assessment. Seismic response reduction device are developed for retrofitting of old facility.
Probabilistic Safety Assessment for AHWR has been performed for Level-1 (internal, Seismic initiating events) and Level-2 (internal events) for estimation of Core Damage Frequency and Large Early Release Frequency. Probabilistic Safety Assessment for critical facilities such as Special Oxide Metal Plant, Waste Management plant, Salt preparation facility for MSBR, etc. was performed to assess the hazardous / toxic scenarios.
Containment studies facility is a 1:250 scale model of standard 220 MWe reactor containment. It has used extensively for containment thermal hydraulic studies during loss of coolant accident. The transient pressure and temperature data post blow down have been evaluated experimentally and used for code validation.
Conducting controlled neutron irradiation experiments is very difficult. To supplement the experiments,a methodology for multi-scale simulation of effects of neutron irradiation on material stress-strain curve and ductility of FCC materials has been developed in this work and its validity to some materials like SS316L has been demonstrated.
An improved welding procedure and technique for long life of the DMW weld joints have been devolved. All the joints have been evaluated based on its weld-residual stress, metallurgical features, mechanical properties and fracture toughness properties of different zones of DMW.
3D detailed CFD computations have been carried out for full transient of hydrogen release during a severe accident in actual reactor containment. 3D full scale geometry of Kaiga 220 Mwe, TAPS 540 MWe, KAPP 700 MWe have been built and used for analysis. For hydrogen mitigation indigenously developed passive catalytic recombiner devices (PCRD) have been tested and qualified in Hydrogen Recombiner Test facility (HRTF), Tarapur.
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The accident source term code PRABHAVINI v2.0 is released in March 2020. The released version 2.0 has five major modules namely PRAVAH (thermal-hydraulics), ABHA (Core Degradation), PARIVAHAN (fission-product transport), PARIRODHAN (Containment phenomena) and PURAK (system components). ABHA contains sub-modules SHAKTI (Point Kinetics), PARIRODHAN contains two sub-modules VAHITI (Containment thermal-hydraulics) and DAHAN (hydrogen combustion).
Facility is built to study the channel disassembly process for PHWR under severe accident heat up conditions. The facility consists of 60 kW power supply, jacket cooling system supported by chiller, Argon gas distribution system. The facility can accommodate scaled down channels. The facility is supported with 60 channel data logger system.
Thermo-mechanical fatigue tests, i.e. under simultaneous, independently controlled cycling of mechanical loads and temperature can be conducted. In addition, high temperature tensile, isothermal LCF and short duration creep tests can also be conducted.
The purpose of the facility is to study various severe accident phenomena related to thermal hydraulics. The facility consists of 36 kW DC rectifier, 20 kW induction furnace, 10 kW infrared heating unit, argon bank and super heater and cooling units.
The purpose of the facility is to study molten fuel coolant interaction, pipe crack flow assessment and PHWR channel blowdown. The facility consists of 50 kW Induction furnace to generate melt at 2400 deg C, 1.25 m3 pressurised hot water generator, circulating canned motor pumps, 2 Te load provider, cooling tower etc.
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The purpose of the facility is to study mainly on PHWR Channel heat up phenomena. This facility consists of 490 kW DC rectifier, 40 kW AC power supply, 10 Te Chiller, data logging station, electrical steam generator and circulating pumps.
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The purpose of the facility is to study severe accident management validation experiments. The facility consists of 60 kW DC rectifier, steam generator and pumps.
The main objective of this project is to undertake collaborative R&D studies on aerosol behaviour under severe accident conditions in the context of Indian Nuclear Reactors by setting up of a state-of-the-art National Aerosol Facility (NAF).
Mechanical testing machine with 10KN (max.) loading capacity capable of testing different types of specimen in the temperature range of RT to 1100 deg.C. It can be used to perform creep, compression, fracture tests at normal and high strain rates.
Biaxial Creep Testing System (BCTS) with a maximum loading capacity of 50 KN can be used to perform biaxial creep tests upto 1000 deg.C and upto 10 MPa internal pressure. Various types of tensile, compression and fracture specimen tests can be conducted.
Universal Testing Machine(UTM) is used to characterize the properties of steel like Young’s modulus, Ultimate strength, yield stress, ductility etc. It is used to evaluate the tensile capacity of various specimens. Plate specimen and round bar specimens can be tested in the machine. The machine is located at ground floor, Hall 11. The capacity of the machine is 1200kN.
Compression Testing machine is used to evaluate the maximum compressive load capacity of concrete cubes or cylinders. It is mainly used to assess the grade of concrete. It can also be used to evaluate the Young’s modulus of concrete cylinder. The machine is located at ground floor, Hall 11. The capacity of the machine is 3000kN.
A state-of-the-art material testing facility for conducting Creep, Fatigue, Fracture and mechanical tests upto 1200 deg.C in a high vacuum environment has recently been installed and commissioned at RSD, BARC. Special fixtures are designed for conducting various high temperature tests on graphite, ceramics and refractory alloys.
The major objectives of this facility are:
\r\n(i) Direct contact condensation phenomenon related to reactor safety systems
\r\n(ii) DCC induced flow dynamics and water hammer
\r\n(iii) Study of bubble dynamics
\r\n(iv)Validate computer code for CIWH phenomena investigation
\r\n(v) Efficacy of spray systems
\r\n(vi) Suppression pool hydrodynamic studies
HRTF facility is a large 60m3 steel vessel constructed to test the performance of different hydrogen recombiner designs developed by TPD, MPD & CD of BARC under dry and wet conditions.
BARCOM is a 1: 64 volumetric scale down model of 540 MWe PHWR containment. Objective of BARCOM containment models was evaluation of structural integrity and containment failure modes for PHWR containment under extreme conditions.
\r\nFollowing tests were conducted in BARCOM containment
\r\n(i) Over pressure Test (OPT) up to the Functional Failure of model
\r\n(ii) Leakage Characterization tests- Local Leak Rate test
\r\n(iii) Proof Test (PT) and integrated leakage rate test (ILRT) up to 1.44 kg/cm2.
The CSF concrete containment is approximately 1:200 volumetrically scaled down model of the 220 MWe PHWR containment of Kaiga Atomic Power Plant. The CSF consists of a Primary Heat Transport System, a Containment System model and a control and instrumentation room. The experiments performed in this facility will generate a large database for the validation of computer codes for containment thermal hydraulics, aerosol and hydrogen transport.
The test set up controls dissolved oxygen levels (>20 ppb) and measures electrolytic conductivity and pH values of demineralised water in low pressure and low temperature recirculation loop. This water is fed to high-pressure (max. 200 bars), high-temperature (max. 350 degree C) autoclave, where cracked specimen is subjected to fatigue cycling.
\r\nThe set is aimed at determination of FCGR constants for realistic assessment of fatigue crack growth life of PWR/ AHWR components under reactor coolant conditions.
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This machine has two independent actuators in axial and torsion directions. Pure axial, pure torsion and combined axial-torsion tests can be conducted at room temperature. In addition, an internal pressure can also be applied in tube type specimens to investigate synergistic damage under ratcheting-fatigue loading.
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\r\nThis machine is aimed at improved assessment of fatigue damage under multi-axial fatigue cycling.
To study the performance of passive cooling of steel containment with water + air and with air and Structural performance of steel containment under accident temperature.
Facilities have been designed, developed and are in operation to carry out (i) thermal ageing (under accelerated conditions up to 300deg Cel), (ii) Humidity Chamber - temperature range from ambient to 80 deg Cel and maximum relative humidity up to 95±5% (iii) radiation ageing - dose rate 0.1MRad/hr , (iv) Loss of Coolant Accident (LOCA) environment - Maximum steam temperature and pressure achievable inside the LOCA chamber are 150 deg Cel and 3.4 kg(g) (50 psig) respectively (v) Main Steam Line Break qualification studies of C&I components and equipment - 171deg C and 32 psig.
Facilities have been designed, developed and are in operation to carry out (i) thermal ageing (under accelerated conditions up to 300deg Cel), (ii) Humidity Chamber - temperature range from ambient to 80 deg Cel and maximum relative humidity up to 95±5% (iii) radiation ageing - dose rate 0.1MRad/hr , (iv) Loss of Coolant Accident (LOCA) environment - Maximum steam temperature and pressure achievable inside the LOCA chamber are 150 deg Cel and 3.4 kg(g) (50 psig) respectively (v) Main Steam Line Break qualification studies of C&I components and equipment - 171deg C and 32 psig.
Various computing facilities/ software tools are available at Engineering Hall No. 3 to perform the risk and reliability analysis of industrial facilities/ complex systems.
Mechanical testing machine with a max 5KN static+ 2KN dynamic loading capacity with a temperature range of -150 to 500 deg.C. Used for small scale specimen tests like small punch tests (SPT), 3-point bend tests, mini-tensile tests, mini-CT fracture tests, mini-fatigue tests. Located at CFB.
A passive Hydrogen Recombiner device has been successfully developed in collaboration with CD, TPD and MPD of BARC. The catalyst designed and developed by CD has been selected for final product development based on performance during qualification experiments. The technology has been transferred to ECIL for large scale production and subsequent deployment in nuclear reactor containments. This PCRD is being deployed in all the existing and upcoming PHWR nuclear reactor containment for hydrogen mitigation purposes.
Diagnostic system is artificial neural networks (ANNs) based operator support system for identification of accident scenarios in 220 MWe Indian PHWRs. An ANN model has been developed to identify various break scenarios of large break LOCA and Main Steam Line Break scenarios in PHWRs. The system has been successfully implemented and demonstrated on high speed computing facility for real-time diagnosis. The source term module for displaying the release of fission products has also been developed and integrated.
Core Safety Studies: Severe Accident Analysis of Indian NPPs for Source Term and Hydrogen
Material Modeling, Simulations and experiments to get properties of aged materials
Strctural integrity assements, fracture, fatigue, creep experiments
Probabilistic Safey Studies: LOCA and Main Steam Line Break (MSLB) test facility
Containment Safety Studies
