advanced heavy water reactor- some licensing issues · 2012-11-19 · atomic energy regulatory...

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



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



Brief Description OfAHWR And Its Safety

Systems



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



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



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



GDWP

RUPTURE ACCUMULATORDISC

RUPTUREDISCECC HEADER

Emergency Core Cooling System

COREREACTORCAVITY



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



VENT SHAFTS(TYP.)

REACTOR BUILDING



Passive Containment Cooler

GDWP



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



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.



Some Experiments in Support of DesignSome Experiments in Support of Design



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.



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



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



How Safety ObjectivesAre Achieved ?



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



Reactor Characteristics

(a) Negative Void Coefficient Of Reactivity

(b) Low Power Density

(c) Low Stored Heat

(d) Axial Grading Enrichment Of Fuel ForMCHFR



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.



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



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



•Floodable Cavity

•Alternate Heat Sink

•Application Of LBB

•Passive System

•Operator Friendly

Other Factors



How PIEs AreSelected?



•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



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



Methodology ForAnalysis



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



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



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



How AcceptanceCriteria Are Selected?



•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



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



Preliminary ResultsOf PSA



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




10-4200 % Inlet Header Break1.2

10-3Main Steam Line Break1.1

FrequencyTransientSr.No.

1) DECREASE IN COOLANT INVENTORY (LOSS OFCOOLANT ACCIDENTS)



A Sample AnalysisResult



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



� 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



RegulatoryReview

•Approval For Safety Criteria

•Project Design Safety Committee



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



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



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



Consequences

frequency

Category 1

Category 5

Contemplated Criteria



Contemplated Technology Independent Criteria



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



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.



Thank You