advanced heavy water reactor- some licensing issues · 2012-11-19 · atomic energy regulatory...
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Atomic Energy Regulatory Board, India 1AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
New NPP Designs and licensing
Advanced Heavy Water Reactor- Some Licensing Issues
S.K.GuptaATOMIC ENERGY REGULATORY BOARD
INDIA
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Atomic Energy Regulatory Board, India 2AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
PRESENTATION
1. Very Brief Description Of AHWR
2. Safety Objectives
3. Major Challenges in Design
4. How Safety Objectives are Achieved
5. General Approach to Safety Calculations - Sets of PIEs,Methodology, Acceptance Criteria
6. Preliminary Results Of PSA
7. Quality in Design & Safety
8. Contemplated Steps in Licensing Process
9.Technology Independent Criteria
10. Treatment Of Passive System
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Atomic Energy Regulatory Board, India 3AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Brief Description OfAHWR And Its Safety
Systems
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Atomic Energy Regulatory Board, India 4AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
ADVANCED HEAVY WATER REACTOR
The AHWR is a 300MWe boiling light water cooled, heavy water moderated, vertical pressure tubetype reactor designed to produce most of its power from thorium. The core consists of (Th-U233)O2 and(Th-Pu)O2 fuel.
Important Features :
• Reactor physics design tuned for the use ofthorium based fuel with negative voidcoefficient of reactivity.• Advanced coolant channel design featureswith easily replaceable pressure tubes.• Passive systems for core heat removal (underboth normal and shutdown conditions),containment cooling and containmentisolation.• Direct injection of ECCS water into fuelbundle.• Advanced accumulator with fluidic device.• Availability of large inventory of water inthe overhead pool inside containment.• No exclusion zone• Associated Desalination Plant
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Atomic Energy Regulatory Board, India 5AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
S ch e m a tic o f A H W R P rim a ry C ircu it
T A IL P IP E
C A L A N D R IA
C H A N N E LF U E L
S T E A M D R U M
IN L E T F E E D E R
D O W N C O M E R
F E E D W A T E R
S T E A M T O T U R B IN E
IN L E T R IN G H E A D E R
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Atomic Energy Regulatory Board, India 6AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
TAIL PIPE
INLET FEEDERCORE
STEAM
DOWNCOMER
FEED WATER
STEAM TO TURBINE
INLET HEADER
COOLANTCHANNEL
VALVEGOVERNOR
IC INLET HEADER
IC TUBES
IC OUTLET HEADER
IC CONDENSATE
GDWP
IC IC
RETURN VALVE DRUM
Passive Decay Heat Removal System
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Atomic Energy Regulatory Board, India 7AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
GDWP
RUPTURE ACCUMULATORDISC
RUPTUREDISCECC HEADER
Emergency Core Cooling System
COREREACTORCAVITY
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Atomic Energy Regulatory Board, India 8AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
To ECCS line
Nitrogengas
Fluidic Device
Accumulator
Fluidic Device
AHWR Fuel Cluster
Water rod
(TH,U233)MOX(TH,Pu)MOX
(Dy in ZrO2)Displacer region
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Atomic Energy Regulatory Board, India 9AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
VENT SHAFTS(TYP.)
REACTOR BUILDING
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Atomic Energy Regulatory Board, India 10AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
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Atomic Energy Regulatory Board, India 11AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
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Atomic Energy Regulatory Board, India 12AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Passive Containment Cooler
GDWP
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Atomic Energy Regulatory Board, India 13AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Safety Objectives
1. No Activity Beyond Plant Boundary for AllDesign Basis Accidents and ATWS.
2. Hundred Year Reactor Life
3. Operator Friendly
4. No Operator Action Requirement for 3 Days
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Atomic Energy Regulatory Board, India 14AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
MAJOR CHALLENGES IN DESIGN
• To maintain stable flow through core under all conditions,the flow beingtwo phase flow and caused by natural circulation
• Driving force in the natural circulation system being very low, hydraulicresistance in the circuit is to be minimised
• To reduce pressure drop, steam drum could not be provided with aseparator which made the task of limiting carryover and carryunder a difficultone.
• To meet thermal margin criterion (Minimum Critical Heat Flux Ratio,MCHFR) without reducing stability margin
• Removal of heat generating fuel is loss of driving force leading to reductionof flow- a phenomenon which poses great challenge to on-load refuelling.
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Atomic Energy Regulatory Board, India 15AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Some Experiments in Support of DesignSome Experiments in Support of Design
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Atomic Energy Regulatory Board, India 16AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
CRITICAL FACILITY FOR AHWR AND 500 MW(e) PHWR
Objectives:Validation of Physics Simulation Modelsand Nuclear Data
Features• Thermal Neutron Flux (Ave) : 108 n/cm2/sec• Nominal Fission Power: 100 Watts• Core: 330 cm IDx500 cm Height• Variable lattice pitch: 20 cm to 30 cm• Types of cores: Reference Core, AHWR core and PHWR core.
Experiments to be performed• First approach to criticality in all the three types of cores.• Dynamic tests for the shut down device.• Meas. of critical height, level co-efficient of reactivity etc.• Assessing coolant voiding reactivity effect.• Measurement of reaction rates and neutron spectrum.• Neutron flux profile measurement by introducing SPND’s.
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Atomic Energy Regulatory Board, India 17AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Simulates main heat transport and safety systems of AHWRMAIN OBJECTIVE• Simulation of accidental conditions, operational transients and start-upSCALING• Power – to – Volume scaling philosophy
adopted• Pressure, temperature and elevation : 1:1• Volume scaling ratio : 452
0.0 0.5 1.0 1.5 2.0 2.52.0
2.5
3.0
3.5
4.0
4.5
5.0
5.5
6.0
6.5
AHWR ITL
Sing
le c
hann
el fl
ow ra
te -
kg/s
Average channel Power - MW
INTEGRAL TEST LOOP FOR AHWR
Comparison of the steady state natural circulation flow rate in ITL and AHWR
ISOLATIONCONDENSER
STEAM DRUMN2 CYLINDER
ADVANCEDACCUMULATOR
TAIL PIPE
GRAVITY DRIVENWATER POOL
RUPTURE DISC
HEADER
FEEDER
ECCS HEADER
FUEL CHANNEL SIMULATOR
INTEGRAL TEST LOOPINTEGRAL TEST LOOP
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Atomic Energy Regulatory Board, India 18AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
INSTABILITY STUDIES FOR AHWR USING NEUTRON RADIOGRAPHY (NRG)
OBJECTIVES:• Develop flow pattern transition criteria• Measurement of CHF, pressure drop, void
fraction and its distribution using NRG• Evolution of Start-up procedureOperating Parameters:
Pressure : 70 barTemperature : 285 0 CNeutron Flux : 106 to 108 n/cm2s
Typical plot for instability due to boiling inceptionTypical plot for CHF during instability
STEAM DRUM
APSARA REACTOR
Test Section
NEUTRON BEAM
CONDENSER
0 10 20 30 40 50 600.00
0.05
0.10
0.15
0.20
0.25
0.30
∆p m
m W
ater
col
umn
Pres
sure
(bar
) (g)
Void
frac
tion
Time (s)0
5
10
15
20
25
30
35
40
Void fractionPressure
Pressure drop
1500 1750 2000 2250 25000
50
100
150
200
250
300
350
400
pressure drop
∆p m
m w
ater
col
umn
7 mm ID test section
TE-16 TE-17
TE-13
TE-14
Trip occured
Tem
pera
ture
0 C
Time (s)
0
20
40
60
80
100
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Atomic Energy Regulatory Board, India 19AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
How Safety ObjectivesAre Achieved ?
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Atomic Energy Regulatory Board, India 20AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Enabling Features To Achieve Safety Objectives
•Reactor Characteristics
•Large Coolant Inventory
•Low Core Damage Frequency
•Direct Injection Of ECCS Water Into Core
•Floodable Core Cavity
•Alternate Heat Sink
•Application Of Leak Before Break
•Passive Systems
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Atomic Energy Regulatory Board, India 21AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Reactor Characteristics
(a) Negative Void Coefficient Of Reactivity
(b) Low Power Density
(c) Low Stored Heat
(d) Axial Grading Enrichment Of Fuel ForMCHFR
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Atomic Energy Regulatory Board, India 22AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Large Coolant Inventory(a) Very Large Ratio Of CoolantAvailable For Directly Cooling The Coreto the Power Produced
(b) Gravity Driven Water Pool
(c) Hydro-accumulators
(d) Steam Drums
• Comparison with Current Reactors.
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Atomic Energy Regulatory Board, India 23AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Low Core Damage Frequency
•Significantly Less Than The Current Value
•Temperature Below Fuel Failure Limits For All DesignBasis Events and ATWS
•Use Of Passive Systems
• Application Of Leak Before Break For All Primary Pipes
•Diversity and Reliability Of Safety Critical SystemsSignificantly Better
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Atomic Energy Regulatory Board, India 24AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Direct Injection Of ECCS into The Core
•Immediate Availability of ECCS At TheCore
•Distribution Of ECC Along The Length AndCross-Section
•No Bypass Of ECC For Most Of The LOCAs
•Differential Flow Rate Of ECCS ForEffective Usability
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Atomic Energy Regulatory Board, India 25AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
•Floodable Cavity
•Alternate Heat Sink
•Application Of LBB
•Passive System
•Operator Friendly
Other Factors
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Atomic Energy Regulatory Board, India 26AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
How PIEs AreSelected?
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Atomic Energy Regulatory Board, India 27AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
•Postulated Initiating Events from Nuclear Safety for PWRs.•Postulated Initiating Events from AERB Draft Guide onDesign Basis Events for IPHWR .•Postulated Initiating Events from INES User’s ManualPrepared by IAEA and NEA (OECD) for PWR, BWR, PHWRand RBMK.•Postulated Initiating Events from Nuclear Safety for BWRs•DBRs for AHWR
LITERATURE SURVEY FOR PIEs
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Atomic Energy Regulatory Board, India 28AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
PIE IDENTIFICATION METHODOLOGY
DETAILS OF PRESENTATION OF A SYSTEMLITERATURE ON PIES OF IPHWRs,
CANDUs, TAPS-BWR, SGHWR, RBMK,VVER AND AGR
IDENTIFICATION OF A PHENOMENA OR A FUNCTION WHOSEABSENCE OR PRESENCE MAY HURT A SAFETY FUNCTION
IDENTIFY CORRESPONDING PIEs
ALL SYSTEMSCOVERED
PRIORITISATION OF PIEs
Task Force Of Safety Engineers and Design Engineers
NO
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Atomic Energy Regulatory Board, India 29AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Methodology ForAnalysis
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Atomic Energy Regulatory Board, India 30AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Step 3: DETAILED THERMO-MECHNICALANALYSIS OF REACTOR CHANNEL
Code Used : HT/MOD4
Step 1: GLOBAL ANALYSIS FOR LOCA OF NUCLEARPOWER PLANT
Code Used:RELAP5/MOD 3.2.2
Step 2: SLAVE CHANNEL ANALYSISCode Used:RELAP5/MOD 3.2.2
Step 5: HYDRAULIC LOAD CALCULATIONSCode Used : RELAP - LOAD
Step 6: TRANSIENT SUBCHANNEL ANALYSISCode Used : COBRA IV - I
Step 4: ACTIVITY RELEASE AND TRANSPORT INPHT SYSTEM
Code Used: PHTACT
POWER PLANT GEOMETRYAND OPERATING THERMALHYDRAULIC CONDITIONS INTHE REACTOR
REACTOR INLET HEADER AND SDP, T, FLOW, QUALITYREACTOR POWER AND CHANNELCHARACTERISTICS
CHANNEL INETFLOW, QUALITY, P, T
METHODOLOGY FOR AHWR ANALYSIS
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Atomic Energy Regulatory Board, India 31AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
GLOBAL ANALYSIS
GLOBAL DENSITY TRANSIENT
HOT CHANNEL POWER TRANSIENT
CONVERGE WITH ASSUMED HOT
POWER
FINAL SOLUTION
CHANGE ASSUMED HOT
POWER
NO
YES
HOT CHANNEL ANALYSIS
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Atomic Energy Regulatory Board, India 32AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
NO EXCLUSION ZONEANTICIPATED TRANSIENT WITHOUT SCRAM AND OTHERBDBAS
ACCIDENT MANAGEMNENTMULTIPLE FAILURE EVENTS
REACTOR STABILITYOPERATING TRANSIENTS
FUEL TEMPERATURE, ENERGY DEPOSITED IN FUELREACTIVITY ANOMALIES
FUEL TEMPERATURESDECREASE IN COOLANT FLOW
REACTIVITY, DESIGN PRESSUREINCREASE IN SYSTEM PRESSURE / DECREASE IN HEATREMOVAL
REACTIVITYINCREASE IN HEAT REMOVAL
SYSTEM PRESSURE, COLD PRESSURISATION,REACTIVITY EFFECTS
INCREASE IN COOLANT INVENTORY
ECCS, FUEL TEMPERATURE AND CONTAINMENTPRESSURE AND TEMPERATURE
DECREASE IN COOLANT INVENTORYDECREASE IN COOLANT INVENTORY
QUALIFYING SYSTEM AND PARAMETERSCATEGORY
CATEGORISATION OF EVENTS FOR AHWR
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Atomic Energy Regulatory Board, India 33AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
How AcceptanceCriteria Are Selected?
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Atomic Energy Regulatory Board, India 34AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
•D.B. NEWLAND, THE DEVELOPING ROLE OF BEST ESTIMATE THERMAL HYDRAULICSCALCULATIONS AND UNCERTAINITY ANALYSIS IN LICENSING IN CANADA, OECD/CSNISEMINAR, ANKARA, 29TH JUNE – 1ST JULY,1998
• IAEA, REVIEW OF FUEL FAILURES IN WATER COOLED REACTORS, TECHNICAL REPORT SERIES388, VIENNA, 1998.
• RBMK CRITERION, LITHUANIAN INTERNATIONAL NUCLEAR SAFETY CENTRE, INGALINASOURCE BOOK,
• FUGEN ATR CRITERION,
• ACCEPTANCE CRITERION FOR ECCS EFFECTIVENESS, DOCKET RM-50-1.
• GUIDELINES FOR ACCIDENT ANALYSIS OF COMMERCIAL NUCLEAR POWER PLANTS APPENDIXB, SPECIFIC INFORMATION RELATED TO PHWRS (CANDU), IAEA, OCTOBER 1999.
• GUIDELINES FOR ACCIDENT ANALYSIS OF WWER NPP, IAEA-EBP-WWER-01, DECEMBER 1999.
• KUDANKULAM PSAR 2001
LITERATURE SURVEY FORACCEPTANCE CRITERIA
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Atomic Energy Regulatory Board, India 35AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
ACCEPTANCE CRITERIA, FUEL CRITERIA AND OTHER CRITERIA
ACCEPTANCE CRITERIA 1 Core coolability criteria for clad temperature to be less than 1200 oC 2 Oxidation criteria of clad surface to be less than 17% 3 Maximum energy deposition in fuel for fuel shattering shall not exceed 200 cal/gm 4 The maximum fuel temperature anywhere in the core shall not exceed UO2 melting
temperature through out the transient 5 Fluid temperature is more than temperature for cold pressurization
FUEL CRITERIA 1 Maximum clad surface temperature shall be Less than 700 oC.(973 K) 2 Maximum energy deposition in fuel for fuel failure shall not exceed 140 cal/gm.
OTHER CRITERIA 1 System Pressure should be less than 110% of Design Pressure
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Atomic Energy Regulatory Board, India 36AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Preliminary ResultsOf PSA
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Atomic Energy Regulatory Board, India 37AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
10-4Double ended break of IC Inlet header1.17
10-2 - 10-3Relief Valve stuck open1.16
10-4Double ended simultaneous rupture of pressure tube andcalandr ia tube.
1.15
10-3 - 10-4Break in the Downcomer1.14
10-3 - 10-4125 mm feedewater line break1.13
10-2 - 10-310 mm instrument line break1.12
10-2 - 10-320 mm Instrument line break.1.11
10-422.2 % ECCS Header breaks with compartmentalisation1.10
10-422.2 % ECCS Header breaks withoutcompartmentalisation
1.9
10-3 - 10-40.352 % Inlet Feeder break with delayed reactor trip1.8
10-3 - 10-40.352 % Inlet Feeder break1.7
10-475 % Inlet header break1.6
10-4100 % Inlet header break1.5
10-450 % Inlet header break1.4
10-420 % Inlet header break1.3
10-4200 % Inlet Header Break1.2
10-3Main Steam Line Break1.1
FrequencyTransientSr.No.
1) DECREASE IN COOLANT INVENTORY (LOSS OFCOOLANT ACCIDENTS)
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Atomic Energy Regulatory Board, India 38AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
A Sample AnalysisResult
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Atomic Energy Regulatory Board, India 39AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
0 200 400 600 800500
510
520
530
540
550
560
570
580
590
600
Tem
pera
ture
(K)
Time (s)0 200 400 600 800
6000000
6500000
7000000
7500000
8000000
8500000
Pres
sure
(Pa)
Time (s)
STATION BLACKOUT WITHOUT SCRAM
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Atomic Energy Regulatory Board, India 40AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
� DBRs Review
� Review Within Safety Analysis Group
� Task - Force 13 Committee
� Design Review Committee
� Internal Review Committee -BARC
� Independent Peer Review - NPCIL
REVIEW AND INTERPRETATION OF RESULTS Internal Review
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Atomic Energy Regulatory Board, India 41AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
RegulatoryReview
•Approval For Safety Criteria
•Project Design Safety Committee
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Atomic Energy Regulatory Board, India 42AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Contemplated Steps InAuthorising Process
Siting Site Evaluation report Design Basis Information Construction Preliminary Safety Analysis Report (PSAR) QA Programme in Design, Construction, and Commissioning PERT Chart for construction schedule
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Atomic Energy Regulatory Board, India 43AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
FSAR; DBMs; Maintenance Procedures (At site)Power Operation
Emergency Preparedness Plan; Complete set of flow and logicdiagram; Commissioning Procedure and test/inspectionprocedure; Pre-requisite commissioning test/inspectionresults; In-service Inspection and Testing Programme;Detailed design description and flow sheets (At Site)
First approach tocriticality
Technical Specification for operation; Training and QualificationProgramme; Training Documents (At site); Operating Manuals(At Site)
FuelLoading/Coolantaddition
PERT Chart for Commissioning ProgrammeCommissioning
Preliminary Safety Analysis Report (PSAR); QA Programmein Design, Construction, and Commissioning; PERT Chart forconstruction schedule
Construction
Site Evaluation report, Design Basis Information Siting
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Atomic Energy Regulatory Board, India 44AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Prelicensing Safety Appraisal•Feasibility of design in principle
•Reasonable assurance of safety as claimed
•Identification of experimental support required
•Identification of standards for variouscomponent performance
•List of PIEs and BDBAs
•Passive system performance
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Atomic Energy Regulatory Board, India 45AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Consequences
frequency
Category 1
Category 5
Contemplated Criteria
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Atomic Energy Regulatory Board, India 46AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Contemplated Technology Independent Criteria
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Atomic Energy Regulatory Board, India 47AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Passive Systems•Reduce the number of components yielding designsimplification
•Derived from operating experience
•More reliable than corresponding active systems
•Eliminate short term operator action
•Reduce dependence on off-site power, movingparts and control system actions for normaloperation
•Degree of passivity
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Atomic Energy Regulatory Board, India 48AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Designers set a safety objective for the designof an Advanced Heavy Water Reactor. Thedesign is supported by several set ofexperiments. Safety studies have been carriedout. AERB working out the process ofauthorising the new design.
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Atomic Energy Regulatory Board, India 49AERB-USNRC Discussion Meeting
29Aug-3 Sept 2004, Washington DC
Thank You