BARC. Technology Development, Design and Safety Features of Indian Advanced Heavy Water Reactor

Transcription

1 BARC Technology Development, Design and Safety Features of Indian Advanced Heavy Water Reactor A.K. Nayak PhD Bhabha Atomic Research Centre Trombay, Mumbai, India IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

2 Design Objectives of AHWR Burning Thorium based fuel Utilization of vast experience of Pressure Tube (PT) design Easily replaceable PT, Improved inspectability and maintainability, Establish fuel performance and material degradation behavior in advance, high reliability fuel handling system Safety enhancement no impact in public domain: Innovative technology Passive safety Competitive capital cost/ MWe & Unit Energy Cost, 1 year design life 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 2

3 Major Design Features GRAVITY DRIVEN WATER POOL (GDWP) STEAM DRUM FUELLING MACHINE CALANDRIA INCLINED FUEL TRANSFER MACHINE AHWR is a vertical pressure tube type, boiling light water cooled and heavy water moderated reactor Important Features Natural circulation heat removal under normal operating and shutdown conditions Low core power density REACTOR BUILDING Power output 3 MWe with 5 m 3 /d of desalinated water. A large fraction (7%) of power from thorium. Extensive deployment of passive safety features 3 days grace period, and no need for planning off-site emergency measures. Design life of 1 years FUEL BUILDING Slightly negative void coefficient of reactivity Direct injection of ECCS water into fuel pins during Loss Of Coolant Accident (LOCA) Easily replaceable coolant channel Large inventory of water at high elevation Passive containment cooling and isolation Utilization of moderator heat Utilization of low grade heat for desalination IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

4 General Arrangement of AHWR IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

5 AHWR Fuel Cluster Key Features Emergency core cooling water injected into the cluster through the holes in displacer tube. Low pressure drop design. IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

6 Concept of AHWR-LEU: Equilibrium core cluster LEU=> 19.75% 235 U and 8.25% 238 U Axial Gradation LEU content (%) in various rings of cluster Inner Middle Outer Cluster Average U-235 Content K ( MWd/Te Xe Sat.) Upper % Lower % % Gd along with (Th,LEU) MOX was used in two pins of innermost ring Fuel pins with Gd Maximum channel power permitted corresponding to ring power factor of 1.15 is 2.85MW IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

7 Physics design features of AHWR-LEU Parameter Values for AHWR-LEU Average discharge burnup, GWd/Te 61 Annual requirement of LEU (net), kg 1167 Energy extracted per ton of equilvalent mined uranium, MWd 7334 Power from thorium, % 38 Peaking Factors (maximum): Local/ Radial/ Axial/ Total Core averaged reactivity coefficients in operating range /1.18/1.5/1.97 Fuel temperature coefficient, k/k/ C: x 1-5 Channel temperature coefficient, k/k/ C: x 1-5 All reactivity coefficients are negative Sufficient reactivity worth of shutdown systems is ensured under all accidental conditions, even with two maximum worth shutoff rods being unavailable Moderator temperature coefficient, k/k/ C: x 1-5 Void coefficient, k/k / % void -8.15x 1-5 Control and safety devices No. of control rods (AR/RR/SR), worth, mk 8 No. each 1.9/11.4/1.3 Total worth of SDS-1 (45 SORs) Total worth of SDS-1 when two rods of maximum worth are not available, mk Physics design ensures inherent safety characteristics of the reactor IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

8 Safety goals for AHWR Reduced Core Damage Frequency (CDF) as compared to the existing plants. Reduced Large Early Release Frequency (LERF)-to an insignificant level Implementation of emergency measures in public domain is not needed Enhance robustness against malevolent acts (insider threats) 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 8

9 FREQUENCY (events/year) Risk Small Reactors Play an important Role in Meeting New Safety Criteria Quantitative Probabilistic Safety Goal unallowable domain - AHWR allowable domain ESBWR Power - Reactor Power Increases Residual risk (RR) : no additional public health concerns - RADIOLOGICAL CONSEQUENCES IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

10 Passive and Inherent Safety Features are Instrumental in Meeting New Safety Criteria The conventional reactors or so called Operating ones have seen an extensive use of active engineering safety systems for reactor control and protection in the past. These systems have certain potential concerning termination of events or accidents that are effectively coped with by a protective system limited by the reliability of the active safety systems or prompt operator actions. Since the reliability of active systems can not be reduced below a threshold and that of the operator s action is debatable, there is growing concern about the safety of such plants due to the large uncertainty involved in Probabilistic Safety Analysis (PSA) particularly in analyzing human faults. In view of this, a desirable goal for the safety characteristics of AHWR is that its primary defense against any serious accidents is achieved through its design features preventing the occurrence of such accidents without depending either on the operator s action or the active systems. That means, the plant can be designed with adequate passive and inherent safety features to provide protection for any event that may lead to a serious accident. Such robustness in design contributes to a significant reduction in the conditional probability of severe accident scenarios arising out of initiating events of internal and external origin. 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 1

11 Example of Applications Passive Systems and Inherent Safety Features in D-I-D in AHWR 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 11

12 Some important passive safety features of AHWR 1/4 Heat removal from core under both normal full power operating condition as well as shutdown condition is by natural circulation of coolant. IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 12

13 Some important passive safety features of AHWR 2/4 Passive Containment Cooling Passive Containment isolation (Th-Pu) MOX Fuel pins Central Tube for ECCS water (Th- 233 U) MOX Fuel pins Passive injection of cooling water, initially from accumulator and later from the overhead GDWP, directly into fuel cluster. IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 13 AHWR FUEL CLUSTER

14 Some important passive safety features of AHWR 3/4 Passive Poison Injection in moderator during overpressure transient Passive Poison Injection System actuates during very low probability event of failure of wired shutdown systems (SDS#1 & SDS#2) and non-availability of Main condenser IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 14

15 Some important passive safety features of AHWR 4/4 Use of moderator as heat sink Water in calandria vault Flooding of reactor cavity following LOCA IAEA Technical Meeting/Workshop December 5-9, , Vienna

16 Passive Systems in Defense-In-Depth of AHWR Level 1 : Elimination of the hazard of loss of coolant flow: Heat removal from the core under both normal full power operating condition as well as shutdown condition is by natural circulation of coolant. Reduction of the extent of overpower transient: Slightly negative void co-efficient of reactivity. Low core power density. Negative fuel temperature coefficient of reactivity. Low excess reactivity 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 16

17 Passive Systems in Defense-In-Depth of AHWR (Contd.) Level 2: Control of abnormal operation and detection of failure An increased reliability of the control system achieved with the use of high reliability digital control using advanced information technology. Increased operator reliability achieved with the use of advanced displays and diagnostics using artificial intelligence and expert systems. Large coolant inventory in the main coolant system. Level 3: Control of accidents within the design basis Increased reliability of the ECC system, achieved through passive injection of cooling water directly into a fuel cluster through four independent parallel trains. Increased reliability of a shutdown, achieved by providing two independent shutdown systems. Further enhanced reliability of the shutdown, achieved by providing a passive shutdown device Increased reliability of decay heat removal, achieved through a passive decay heat removal system, which transfers the decay heat to GDWP by natural circulation. Large inventory of water inside the containment (about 6 m3 of water in the GDWP) provides a prolonged core cooling meeting the requirement of grace period. 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 17

18 Passive Systems in Defense-In-Depth of AHWR (Contd.) Level 4: Control of severe plant conditions, including prevention of accident progression and mitigation of consequences of severe accidents Use of moderator as heat sink Flooding of reactor cavity following a LOCA Level 5: Mitigation of radiological consequences of significant release of radioactive materials The following features help in passively bringing down the containment pressure and eliminates any releases from the containment : Double containment Passive containment isolation Vapour suppression in GDWP Passive containment cooling 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 18

19 Peak Clad Temp v/s frequency of occurrence a quantitative probabilistic safety criteria Temperature( C) Large Break LOCA without ECCS BDBEs Decrease in coolant inventory Increase in coolant inventory Increase in heat removal Increase in system pressure/decrease in heat removal Decrease in coolant flow Reactivity anamolies Operational occurances/transients Multiple failure events Wires system failure events 2 % LOCA DBEs AOO & NO 2 1E-111E-1 1E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E Frequency 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 19

20 Core Damage Frequency Per Year AHWR ~ 1x1-8 Ref: Lecture on Near Term Advanced Nuclear Reactors and Related MIT Research, by Prof. Jacopo Buongiorno, MIT, USA, June 16, December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 2

21 PLDSC review of AHWR Some of the passive and inherent safety features are already present in operating reactors; however some are new, for example Natural circulation heat removal at rated condition for a boiling pressure tube type reactor; Passive ECCS injection directly into all the channels during a LOCA; Decay Heat Removal Using ICs; Passive Containment Cooling and Containment Isolation; Passive Shutdown Reduced flexibility of passive systems in abnormal conditions Start-up procedure from cold start-up is very long. For a First-of-a-kind Design, it is essential to establish their capability and reliability, and hence ability to meet the safety goals. 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 21

22 PLDSC review of AHWR Intended goal 1 years of life; no proven technology for existing reactors. Fluid status in water tubes and ECCS header? Effect on reactor physics; corrosion issues for stagnant water; blockage of holes due to corrosion products; detection of blockage during operation? Thermal stratification in GDWP Natural circulation behavior during refueling Stagnation channel break; possibility of whole core melt down and management of severe accident? 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 22

23 What has been Accomplished so far? Experimental facilities have been used to gain insight of singlephase and two-phase natural circulation in single and multi channel configurations over a wide range of pressure and for validation of inhouse codes Steady State Performance experiments were carried out at different powers in ITL. Stability Performance experiments have been performed at various powers in ITL and compared to in-house code TINFLO-S results. The reactor start-up has been demonstrated experimentally in ITL. It has been found that the flow instabilities during start-up are encountered only below system pressure of 35 bar. CHF related experiments carried out in 3 MW BWL and Freon loop at IIT, Bombay. Experiments revealed that adequate margin exists in AHWR fuel bundle under 12% full power. 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 23

24 Experimental Facilities for AHWR Design Validation BARC Facilities AHWR Critical Facility Integral Test Loop (ITL) Natural Circulation Loop (NCL) High Pressure Natural Circulation Loop (HPNCL) Flow Pattern Transition Instability Loop (FPTIL) Parallel Channel Loop (PCL) Passive Containment Isolation Test Facility (PCITF) Passive containment cooling test facility Passive concrete cooling test facility Boiling Water Loop (BWL) Isolation Condenser Test Facility (ICTF) Passive External Condensation Test Facility (PECTF) Air Water Loop (AWL) AHWR Thermal-Hydraulic Test Facility (ATTF) Fuelling Machine Test Facility (FMTF) Containment Systems Integral Simulation Test Facility (ConSIST) IIT Bombay Thermal Hydraulic Test Facility (THTF) Freon Loop for CHF studies Passive Moderator Cooling System IIT Kharagpur Facility for carry over and carry under studies Facility for Re-wetting experiments Development Activities Fuel Rod Simulator (FRS) development (direct & indirect), Power decay and coast-down simulation Two-phase instrumentation (flow, void fraction, channel power) Passive valves (HSPV, AIPV, PPIV) 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 24

25 FCS AA JC SB Integral Test Loop (ITL) Scaling Philosophy GLOBAL SCALING Power to Volume scaling Pressure, temperature &elevation- 1:1 Volume scaling ratio- 452 BOUNDARY FLOW SCALING Feed water and steam flow simulation Pressure, temperature & enthalpy- 1:1 LOCAL PHENOMENA SCALED ARE CHF, Geysering, flashing, Carry-over and carry under in steam drum, etc. Steam Drum IC SD SFP PBC HEADER 54-rod FRCS BFST Start-up boiler Header Isometric of ITL 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 25

26 Pressure - bar Power - kw Feeder Flow - lpm Inlet Temperature - O C Power (kw) Steady state performance of natural circulation in MHTS Main objectives of ITL Mass flow rate Pressure drop void fraction CHF under steady and oscillatory conditions Gravity separation of Steam-water mixture in SD Stability performance of natural circulation in MHTS Static instability Dynamic instability Safety systems Passive decay heat removal system (ICS) ECCS during LOCA Verification of Start-up Procedure and all AOOs Time - s 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna Stable 5 1 Subcooling (K) 15 Experimental Data Unstable data Stable data 2 Unstable 25 8 ITL Cold Start-up at 2% FP and 1 bar Power FCS inlet temperature 6 Flow Pressure (bar) Pressure

27 Mass Flow Rate (kg/s) Steady State and Stability Performance of NC Experiments carried out at different powers in ITL and compared with predictions of various codes. Measured and predicted steady state flow rate are in good agreement. Stability Performance Stability performance predicted by in-house code TINFLO-S. Experiments performed at various powers in ITL Good match between predicted stability surface and test data obtained 3 Pressure = 7 bar T sub = 1-2 K 12 Unstable data Stable data Power (kw) Experimental data 3 Earlier experimental data Analytical model TINFLO-S RELAP5/Mod3.2 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 27 Power (kw) Stable 4 Pressure (bar) Unstable 15 2 Subcooling (K) Fig.16: Comparison of experimental instability data and theoretical p 25

28 Core Flow Rate (kg/s) Temperature ( o C) Density (kg/m 3 ) Pressure (bar) Void Fraction (%) Power (kw), Flow rate (lpm) Steam Drum Pressure (bar) Validation of Start-up Procedure for AHWR At low system pressures flow instabilities are encountered in two-phase natural circulation loops at the boiling inception. This can be avoided by pressurizing the system externally. A stage-wise external pressurization at constant 2 % reactor power has been adopted in AHWR start-up procedure. The reactor start-up has been demonstrated experimentally in ITL. It has been found that the flow instabilities during start-up are encountered only below system pressure of 35 bar Core inlet Density SD Pressure Steady state at 2% FP Pressure controlled SD Pressure Self pressurization Core flow rate core inlet Temperature Core exit Void fraction Time (s) Start-up transient for AHWR (stage-wise external pressurization and boiling inception at 7 bar to avoid low pressure instabilities) Fig.21.11: Predicted start-up transient using RELAP5/MOD3.2 (Scheme - 1 with stage-wise pressurization up to 7 bar) December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna Time (s) Flow FCS Power ITL start-up with external pressurization upto 35 bar

29 IC Pool Temperature - o C SD Pressure - bar IC Pool Level - m ICS Design Validation in ITL Isolation Condensers (ICs) in AHWR ICs are Vertical Heat Exchangers submerged in Gravity Driven Water Pool (GDWP). Steam condenses in vertical pipes of Ics rejecting heat to GDWP. Condensate returns by gravity to steam drum. ICS Performance Tests Isolation Condenser performance validation tests completed in ITL in the range of power 2-6% FP ICS performance is found to be as per design IC initiates Pressure SD Pressure (Expt.) SD Pressure (R5) IC Pool level (Expt.) IC Pool level (R5) Level Initial Pool Level m Time - s thermocouple in at level m (Expt.) thermocouple in at level -.6 m (Expt.) temperature of at level m (R5) temperature of at level -.6 m (R5) Isolation Condenser (IC) of ITL installed on FISBE building Initial Pool Level m Time - s IC test at 75 kw (3% FP) and 1.75 m initial pool level simulated with RELAP5/Mod3.2 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 29

30 ECCS Design Validation in ITL 1 LOCA Simulation Tests Header Pressure (RELAP5) Accumulator Pressure (RELAP5) Header Pressure (Expt.) Accumulator Pressure (Expt.) 9 8 For large breaks (e.g. 21% - 2%), the ECCS injection by Advanced Accumulator (AA) followed by GDWP has been found to be adequate For small breaks (e.g. 9.6 & 5%), after completion of accumulator injection, there was a time gap for GDWP injection to begin Predicted and measured transient behavior matched closely Pressure - bar 7 6 Inlet Header Break - 1% Initial FCS Power - 52 kw ( 2% FP) AA without fluidic device Time - s Variation of header and accumulator pressure following 1% inlet header break in ITL 1.6 Accumulator flow (RELAP5) GDWP flow (RELAP5) Accumulator flow (Expt.) GDWP flow (Expt.) 1.4 Mass flow rate - kg/s LOCA experiments were carried out in ITL for inlet header break sizes varying from 5-2% 1.2 Accumulator GDWP Time - s Variation of accumulator flow rate following 1% inlet header break in ITL IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 3

31 FUEL ROD CLUSTER SIMULATORS FLC-1 & 2 SB AHWR Thermal Hydraulics Test Facility, (ATTF) Tarapur Objective: To evaluate the Thermal margin of the AHWR. To evaluate the Stability Margin of AHWR Power to volume scaling philosophy Pressure, Temp, elevation Scaling: 1:1 Power, Volume Scaling : 1:226 Maximum power to test section: 9 MW Containment Systems Integral Simulation Test Facility (ConSIST): to validate the containment design and performance of its associated systems like GDWP (suppression pool), PCCS and PCIS following a LOCA Fuelling Machine Test Facility IH - Inlet Header ECCS-Inlet Header SB - Startup Boiler FLC- Full Length Channel JC - Jet Condenser SC - Subcooler SFP- Secondary Feed Pump POOL BOILING COOLERS STEAM DRUM JCs SC IH DMWST GDWP ECCSH SFP ATTF Layout BFSTP BFST 7 December 211 IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 31

32 Separate Effect Test Facilities for NC and Other Passive Systems High Pressure Natural Circulation Loop Parallel Channel Loop Passive Containment Isolation Test Facility PCL Passive Containment Cooling Test Facility Boiling Water Loop U-duct with transparent section Test Facility for Moderator Flow and Liquid Poison Distribution FPTIL/CHIL IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 32

33 Heater Average Void Fraction Heater Power (kw) Pressure Drop across Feeder (mm of WC) Components / Sensors Developed Acoustic sensors to detect small breaks Pressure: 5 bar Initial Power: 9kW Void fraction Power delp with feedback Time (s) Void Reactivity Feedback Simulation -2 Channel power measurement FUEL 54-rod CHANNEL FCS SIMULATOR (FCS) Multi-point conductance probe Void Sensors for Steam- Water Systems Two Phase Flow meter GAMMA RAY FRACTION METER MOUNTED ON ROTARY/LINEAR TABLE QOV to simulate LOCA (Opening time: 4 ms) Two Phase Flow Sensors - Single beam Traversing Gamma Ray Densitometer Accumulator AIPV TO ECCS Header IAEA Technical Meeting/Workshop December 5-9, 211, Vienna Passive valves (HSPV, AIPV, PPIV) 33

34 Managing Fukushima Type Accident in AHWR Postulation of Fukushima Type Accident A strong earthquake with/without Tsunami causing prolonged SBO for several days (~ 1 days) Reactor tripped on seismic signal. Steam line is isolated due to closure of CIESV and feed line is unavailable. MHT is boxed up; pressure rises to the set-point of IC. IC is valved in at 76.5 bar by the opening of passive valve. Active valve in IC opens after 3 minutes due to loss of compressed air. Upon MHT depressurizes due to cooling by ICs, Accumulator injection occurs. If MHT pressure falls below the set point of GDWP, injection from GDWP may occur. GDWP water temperature continues to rise, which causes V2 air temperature to increase causing containment pressure to rise. When boiling in GDWP occurs, steam generated condenses in the containment walls and PCCS tubes. Containment venting is not credited for 1 days. IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 34

35 Pressure(bar) Pressure(bar) MHT and containment Pressure Active Valve opens Passive Valve opens Containment Pressure 4 Reactor Trip Accumulator injection starts MHT Pressure Time(days) GDWP water boiling starts Time(days) MHT Pressure rises to set point of ICs within 1 minutes and accumulator injection starts at nearly 45 minutes MHT pressure remains nearly above 3 bar after nearly 2hrs and containment pressure rises to 1.35 bar after 1 days IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 35

36 Temperature( C) Temperature( C) Clad Surface Temperature Clad Surface Temperature GDWP Water temperature Time(days) Clad surface temperature remains around 13 deg C even after 1 days of accident 4 Time(days) GDWP water boiling starts after nearly 8 days. GDWP has more than 75m3 even after 1 days. IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 36

37 Steam drum Pressure (bar) Reactivity (mk) PPIS is effective in handling failure of wired shut down system AHWR has an unstable operational regime at low power and high subcooling conditions since it is a natural circulation BWR. Operation in these regimes are forbidden and if it does, reactor is tripped. However, if the wired shut down systems is bypassed due to some malevolent action and all the RRs are withdrawn, a maximum of 11 mk is added, With addition of the reactivity, the reactor power increases. When the steam flow rate reaches more than 11 %, MSIV closes. Primary side pressure increases to the set point of PPIS (83 bar) enabling shut down of the reactor passively. Before that, IC is valved-in at a pressure of 76.5 bar to 79,5 bar due to opening of passive valves and then by opening of active valve at 8 bar. Decay heat is removed passively by Ics Time (s) PPIS actuation at 83 bar 9 PPIS actuates SD Pressure Time (s) IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

38 Reactor Power (MW) Clad Surface Temperature (K) Performance during a LORA Unstable operation at 3 % FP Reactor Power PPIS actuation Clad Surface Temperature Time (s) LORA at s Reactor Power Time (s) Reactor power increases to maximum 11 MW and the peak clad surface temperature is 585 K IAEA Technical Meeting/Workshop December 5-9, 211, Vienna

39 Subcooling (K) % Full power Water level in tank (% of design value) Failure region Passive system reliability A methodology APSRA has been developed indigenously to assess the passive system reliability. APSRA is being considered among the state-of-art methods internationally. Existing methods like REPAS and RMPS are based on propagation of uncertainty in the parameters through BE codes using some probability density function. In view of arbitrariness associated with treatment of critical parameters, APSRA relies on generation of a failure surface and the root diagnosis to attribute the causes of deviation of critical parameters. APSRA relies on comparison of test data with BE predictions to treat modeling uncertainty. APSRA has been applied to the passive systems of AHWR like NC in Main Heat Transport System, Isolation condenser system etc Constant % full power lines Pressure (bar) Failure probability for Natural Circulation ~ 3x1e-9/ yr Failure frequency 3.5E-9 3E-9 2.5E-9 2E-9 1.5E-9 1E-9 5E-1-5E IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 7 Pressure (bar) Subcooling (K) Boiling Natural Circulation in MHTS of AHWR Failure due to insufficient V1-V2 pressure differential to raise water to spill into duct Success region Failure due to insufficient inventory in the tank to form liquid seal Failure region % Break size Failure surface for Passive Containment Isolation System 39

40 Summary Several passive features have been adopted in AHWR design to enhance the safety and to eliminate any impact in public domain. Both theoretical and experimental validation successfully completed for many design aspects of the AHWR. Additional experimental validations are planned in BARC and ATTF Tarapur for the remaining issues. PLDSC review of AHWR has been completed. Site evaluation is in progress. IAEA Technical Meeting/Workshop December 5-9, 211, Vienna 4