PHWR Safety Review Experience - Nuclear Regulatory Commission · 2012-11-19 · Atomic Energy...

1Atomic Energy Regulatory Board, India AERB-USNRC meeting30 Aug - 3 Sept 2004, Washington DC

PHWR Safety Review Experience

Experience With Design Safety Review OfIndian Pressurised Heavy Water Reactors

Some Examples

R.I. Gujrathi, S.A. Bhardwaj



Reactor Shut Down Systems


Reactor Shut Down Systems

RAPS & MAPS units have moderator dumping as the reactor shut downsystem.

In place of moderator dumping, two equally fast acting shut downsystems namely

• Primary Shut down System (PSS) / Shut Down System-1• and Secondary Shut down System (SSS) / Shut Down

System-2 based on diverse principles have been adopted for all

subsequent PHWRs in India.


RAPS/ MAPSRAPS/ MAPSModerator dumping asModerator dumping as

S/D systemS/D system


INTEGRAL ASSY. OF CALANDRIA & END SHIELDS


Primary Shut down System(PSS)• Cd sandwiched tubular

elements at 14 locationsin 220 MWe units.

• Drops into moderatorvolume on release of anelectromagnetic clutch.


CALANDRIACALANDRIA

Helium GasHelium GasTankTank

PoisonPoisonAdditionAddition

TankTank

PoisonPoisonPreparationPreparationand Drainand Drain

TankTank

LiquidLiquidPoisonPoisonTubesTubes

From cover gas systemFrom cover gas systemTo cover gas systemTo cover gas system

Fast-acting solenoid valvesFast-acting solenoid valves

SECONDARYSHUTDOWN SYSTEM-

220 MWe units ( 4 Banks of3 tubes each)

POISONPOISONTANKTANK


The design and subsequent testing anddevelopment of these new systems were followedand reviewed by expert committees of AERB,particularly with respect to demonstration of reliableoperation.

(The tests conducted on prototype, life cycle testingfor an actual assembly , results of tests on allassemblies during commissioning are reviewed.Any subsequent change in design of system orcomponents is permitted by AERB after properqualification process)


• During testing and commissioning deficiencieswere observed and many improvements wereproposed by AERB review teams. All suchimprovements could be incorporated in thesystems.

• On power clutch testability in PSS was not builtin the original design. This feature has beenincorporated for later units


LOCATION OF CONTROL AND PROTECTION DEVICES


SHUT- DOWN SYSTEMS – Reactivity worth Given the limitations

� of available space in a small core of 220 MWe units� PHWR philosophy of keeping regulating devices and

protective devices independent of each other� and using the design developed for SSSthe number of units of these fast acting PSS rods and SSSbanks which could be located on calandria were limited. Thereactivity worth in individual systems was sufficient toadequately ‘make’ the reactor sub-critical taking into accountthe gain in reactivity due to:

� Changes in Fuel temperature / Power on shut down� Cool down of coolant and moderator� Void in coolant during LOCA� Loss of regulation events but not ‘hold’ the reactor sub-critical for prolonged period


The reactivity depth provided individually by PSS or SSS could notaccommodate large XENON swing following a reactor shut down.

Though XENON decay following a reactor shut down is a delayed and aconsiderably slow transient , AERB directed this part of slow reactivityaddition also to be compensated by automatic system and not by manualaction.

NPCIL proposed Automatic Liquid Poison Addition System, ALPAS, for thispurpose. This system, also referred as Bulk Addition Mode (ALPAS-BAM)injects Boron solution into moderator flow, before it enters the Calandriavessel.

ALPAS Controlled addition Mode ( ALPAS -CAM), a feature separatelyprovided, supplemented the worth of reactor regulating system, if needarises. Reactor regulation in normal mode is provided by solid rods(Cobalt or stainless steel).

Thus (PSS + ALPAS) OR (SSS + ALPAS) PROVIDE ADEQUATE Shut downmargin under all design basis events.


However ALPAS depends on availability ofmoderator flow and thus is ineffective underSBO.

AERB desired ALPAS action to be madeindependent of availability of moderator flow.


Gravity addition of boron(GRAB), a systemincorporating manualaddition of boron into themoderator by gravityduring SBO was acceptedfor NAPS & KAPSreactors

Liquid Poison InjectionSystem- LPIS; a newsystem was developedand incorporated in laterunits, to make directinjection of boron poisoninto moderator possibleautomatically –usinghelium pressure

TANK CONTAINING B2O3SOLUTION

LPIS is identical system withpoison tank getting pressurised toinject poison through automatedopening of valves

GRAB


• SSS can possibly drainback after actuation if Hefrom gas tank leaks

• SSS system has 4 totallyindependent banks, acommon leak in all Hetanks ruled out

• demonstrated to be leaktight duringcommissioning

• ALPAS amount and logicof actuation was soselected as to minimizemission time for SSS.

• System for 540 MWeunits redesigned toovercome this.


SHUT DOWN SYSTEM # 2 ( 540 MWe units)


Shut Down System # 2Shut Down System # 2Single Jet Experiment


Shut Down System # 2Shut Down System # 2Multiple Jet Experiment


Shut Down System # 2Shut Down System # 2System Actuation Test


� 540 MWe PHWRs, Large Core

� Shut- down System 1:28 mechanical Shut-off Rods of Cd

� Shut-down System 2:

� Designed to inject Gd Nitrate soln in moderator using He pressure

� Poison mixes with moderator unlike in 220 MWe PHWRs, whereinpoison is injected in tubes

� Adequate reactivity worth of each of the shut-down systems forlong-term sub-criticality

� Reactor Regulating System comprises of Liquid Zone Controlsystem using light water as absorber

TAPP-3&4: Shut down & Regulating Systems


Liquid Zone ControlLiquid Zone ControlSystemSystemTest Loop


Liquid Zone Control SystemLiquid Zone Control SystemTest Loop


Critical Measurements

•Worth of shut down systems - demonstration ofpower reducing capability

•Power run down test( dynamic testing of SDSs - 14 PSS, 13 PSS, 4SSS BANKS , 3 SSS BANKS, ALPAS-BAM,GRAB,LPIS)

•Calibration of control devices and ALPAS-CAM (mandatory for all power reactors)


Critical Measurements

• Determination of static worth of shut down systems

critical experiments with and without shut down system (14 PSS, 13PSS, 4 SSS banks, 3 SSS banks) with various configuration of controldevices

Such experiments are repeated whenever

• Initial core loading is different from the standard core.

• Initial core loading with 384 depleted uranium bundles forpower flattening and rest natural uranium bundles

• Initial core loading with Thoria bundles for power flatteningand rest natural uranium oxide bundles


FU EL LO ADING INITIAL FO R K APS-1

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 A A

B B

C C

D D

E E

F 3 F

G 10 6 10 G

H 2 7 10 H

J 3 3 9 3 J

K 7 6 8 6 3 3 K

L 9 10 5 7 3 9 L

M 2 5 4 11 M

N 7 6 N

O 2 10 3 3 OP 3 PQ QR RS ST 10 T

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20

ADJUSTERS ADJUSTERS


0 20 40 60 80 100 120 140 160Full Power Days

6.2

6.4

6.6

6.8

7

7.2

7.4

7.6R

eact

ivity

(mk)

31

32

33

34

35

36

37

38

Reactivity (m

k)

SR14PSS13PSS+SR

WORTH OF SHUTDOWN DEVICES


0 20 40 60 80 100 120 140 160Full Power Days

6.2

6.4

6.6

6.8

7

7.2

7.4

7.6R

eact

ivity

(mk)

27

28

29

30

31

32

33

34

35

Reactivity (m

k)

SR4SSS3SSS+SR

WORTH OF SHUTDOWN DEVICES


31.533.96 ± 2.014 PSS rods1

24.826.7 ± 2.013 PSS rods2

23.222.6 ± 2.03 SSS banks3

CalculatedMeasured

Reactivity Worth in NAPS1 core(mk)

ReactivityDevicesNo


Sub Critical Measurements

To verify the absorption characteristics of individual devices ofreactor regulating system (RRS) and protection system (RPS)

For RPS - individual rod Individual bank

13 PSS rods3 SSS banks


0 2 4 6 8 10 12 14 16 18 20TIME (SECS)

-35

-30

-25

-20

-15

-10

-5

0

REA

CTI

VITY

(mk) ION CHAMBER

CH-DCH-ECH-J

14 PSS RODS DYNAMIC TEST


9.259.22BANK B1 IN

9.0610.0BANK B2 IN

8.528.66BANK B3 IN

9.659.08BANK B4 IN

CALCULATEDMEASURED

WORTH IN MK

PSS STATUS

Worth of Individual SSS Banks by Single Bank Firing and Draining


1.492.01ROD M1 IN

1.672.19ROD M3 IN1.081.52ROD M5 IN1.542.36ROD M7 IN

CALCULATEDMEASURED

WORTH IN MKPSS STATUS

Worth of Individual PSS Rods by Single Rod Insertion

19.1819.7PSS BANK 2

14.617.23PSS BANK 1

36.93 33.79ALL 14 RODS

TOTAL WORTH IN MK


•GRAVITY ASSISTED BORON ADDITION•PROCESS INEPENDENT•DIRECT ADDITION OF BORON TO MODERATOR•NEEDED DURING SBO STATE•BORON GETS ADDED BY DISPERSION

CALANDRIA

TANK CONTAINING B2O3 SOLUTION

�LPIS is identical system with poison tankgetting pressurised to inject poison throughautomated opening of valves

GRAB


0 2000 4000 6000 8000 10000 12000 14000Time (seconds)

-50

-45

-40

-35

-30

-25

-20

-15

-10

-5

0

Rea

ctiv

ity (m

k)

PSS-DPSS-ESSS-GSSS-H

LPIS WITHOUT MODERATOR CIRCULATION


0 200 400 600 800 1000 1200 1400TIME (SECS)

-140

-130

-120

-110

-100

-90

-80

-70

-60

-50

-40

-30

-20

-10

0

REA

CTI

VITY

(mK

)

PSS-DPSS-ESSS-GSSS-H

LPIS WITH MODERATOR CIRCULATION



Emergency Core Cooling System



ECCS in early PHWRs in India (RAPS / MAPS)

• Low pressure injection by moderator pumps• Low pressure long term recirculation of spilled water• Back up supply by fire water system



AERB desired strengthening of ECCS (PARTICULARLYAFTER TMI-2) BY

• Incorporating high pressure Injection system� Analysis and provision of handling variety of break sizes

including small leaks• Testing of system during commissioning



Accordingly NPCIL designed ECCS to act in three stages.

• High pressure heavy water injection

• Medium pressure light water injection

• Low pressure long term recirculation via suppression pool


Emergency Core Cooling SystemCommissioning tests revealed ambiguity in

• Level measurement in accumulator during the process of injection• ?P measurement across PHT headers for selection of typeof injection (type-I, type-II or type-III)

These were overcome by redesign and are further simplifiedin the present 540 MWe design.( Prior to light water draining, the test is witnessed andcertified by regulatory body)



• High pressure heavy water injection system hasalso been back fitted in RAPS & MAPS.

• However, AERB has accepted shared systemapproach for this back fitting; with LOCA occurringin one unit, the other healthy unit will be manuallyshutdown and brought to cold shut down state.


Experience With Design Safety ReviewOf

Indian Pressurised Heavy WaterReactors

Containment


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Evolution of PHWRsCotainmentdesigns inIndia

1


Containment: Design Improvements

In Kaiga-1 To 4 And RAPP-3 To 6

� Many improvements were based on outcome of thedetailed design review of NARORA andKAKRAPAR plants

� Primary containment is dome shaped instead of flatroof

� Steam generators within the primary containment� Separate structural wall, vent shaft, wall surfaces

accessible for inspection and painting� Secondary containment blow out panels eliminated� Guard pipes for all high enthalpy lines in Secondary

containment.


NAPS REACTOR BUILDING


REACTOR BUILDING: KGS-3,4 AND RAPP-5,6


Containment Systems Air Ingress In The Boxed-up Containment

� Primary Containment (PC) pressure followingLOCA could rise due to compressed air ingress inthe boxed-up containment.

AERB review required to minimise air ingress inthe boxed-up containment


Separate header to supply compressedair to essential valves and instrumentsfor accident management provided ( alsoback fitted in old plants).

This change has reduced air ingress rate fromabout 4.5 m3/min about 1.5 m3/min in NAPS typecontainments resulting in reduction in estimateddose at the exclusion boundary.


Primary Containment Filtration And Pump Back System

� Combined HEPA + charcoal filters� Reduction in concentration of radio nuclides� Modifications to facilitate direct measurement of

charcoal bed temp in place of outlet air temp� Charcoal temp maintained below the desorption

temperature� Thermal analysis of charcoal beds conducted� System can be brought in service after 4 hours

following accident – decay of I - 134� Modifications also carried out to improve reliability

of the system


Figure


Air Locks

� Nitrogen gas supply, as a back-up, to inflate door-seals of air-locks during station black out - tomaintain the containment integrity under suchsituations


Failure, Causes Of FailureAnd Re-engineering

OfKAIGA IC Dome


Indian PHWR Containment� Double containment

• Pre-stressed inner containment (IC)

• Reinforced concrete outer containment (OC)

� Unique feature of IC dome

• Four large openings for removal and replacement ofsteam generator

• These are asymmetric about N-S and symmetric aboutE-W of containment structure


Delamination Of Kaiga-1 IC Dome

• Design report for kaiga-1 IC JAN 93 domereceived by AREB

• Preparatory work for JUL 93 constructionof dome commenced

• AERB approval NOV 93

• Casting of dome commenced DEC 93

• Pre-stressing work started APR 94

• Delamination incident MAY 94


On 13/5/94 in kaiga-1, someportion of IC dome delaminated

Undersurface of the domecollapsed 7 hours after stressingof 66 cables out of total 183cables

At the time of the incident, thework of removal of steelsections of the central derrick,used for supporting the domeshuttering, using the hooksprovided in the surface of thedome was in progress

Damaged surface wassymmetrical about E-W centralline


• Cables which had been stressed lost fully/partiallythe stress

• Reinforcing bars pulled out of debris and werehanging

• For investigation

AERB constituted Engineering Committee forInvestigation of Dome Failure (ECIDF).

Failure Incident ..(Contd)


• The underside of the IC dome collapsed due todelamination under the effect of radial forces generated bystressing of the curved cables

• Progressive delamination occurred due to

• Radial tensile stresses causing cracking of concrete.

• Failure of bottom layer of dome surface due to bendingand tearing of the concrete by concentrated force at thecorner of the straightened cable in the delaminated zoneand the curved cable in the remaining portion of thedome. Failure continued till the tension of cable reducedbelow tearing capacity.

Failure Mechanism


The weakness of the dome in withstanding the effect of

• Radial tension combined with membranecompression generated by stressing of cables

• Inadequate provision for resistance to radialforces through radial reinforcement connectingthe top and bottom concrete layers at criticalzones due to congestion

• Analysis results showed max. radial stress tooccur in the zone where highest depth ofdelamination occurred; in this zone cables wereclosely spaced

Causes Of Dome Failure As Brought By ECIDF (AERB)Committee


Based on the recommendation of ECIDF (AERB)committee

• IC dome thickness has been increased to 470mm from 340mm

• Concrete grade has been changed from M45 to M60

• Cable spacing has been changed to 225 mm

• Thickness around SG opening changed to 1200 mm from1000 mm

• Transition taper at SG opening made more gradual from1:3 to 1:4

Re-engineering Of IC Dome


• Radial reinforcement is provided

• Programmatic improvement

• Strict implementation of Q.A. Plans

• Mockups carried out for complicatedareas before actual construction

Improvement In Dome Design



Annulus Gas System


Annulus Gas System (AGS)• Annulus space between coolant tube and calandria tube

of earlier stations is open to calandria vault. This causesproduction of Argon-41 activity.

• AERB desired means to minimise Argon-41 in newplants.


• To overcome Argon-41 problem calandria vault in NAPS onwards is filled withwater and annulus space between coolant tube and calandria tube was also filledwith CO2. Based on PICKERING experience of pressure tubes failure, AERB desired tointroduce monitoring of pressure tube for any leak of Coolant tube. Coolant tube leak detection by AGS was implemented. . An important function vital to safety played by AGS is of detection andidentification of a leak in the coolant channel/calandria tube by monitoring the dewpoint of annulus gas (CO2) which is circulated through the annular space, betweencoolant tube and calandria tube


In NAPS & KAPS AGS system is operating on purge mode where CO2from cylinders is purged through annulus space and exhausted to stackthrough ventilation exhaust.In this case also Argon-41 problem is observed occasionally, if systempressure falls due to errors at time of replenishment of CO2 cylinders.AERB desired to solve the problem. This has been overcome byintroducing recirculation mode operation and keeping the system in positivepressure to avoid ingress of air into the system.Two blowers (1 operating + 1 manual standby) are provided in KAIGA-1,2& RAPS-3,4 onwards for the operation of the AGS on recirculation-mode.

Based on international experiences:-

- system is improved to identify leaky coolant tube / string in availabletime to satisfy leak before break criterion.- to maintain oxide layer on pressure tube and to minimise D2 ingressin pressure tube, oxygen addition in CO2 is introduced.- beetle alarm, blower’s discharge pressure high alarm and hightemperature annunciation are also provided for indication of leak into AGS.



Process Water


Process Water System� RAPS-1&2:� PW Is used to reject heat from moderator in an open

loop to the lake.� Activity discharge to the lake

� MAPS:� Intermediate closed-loop PW system� NAPS:� AERB review suggested to provide reliable back-up water

supply to PW supply as a design improvement.� Fire-water back-up provided for Moderator HX, S/D Cooling

HX, ECCS HX etc. (End-Shield cooling sys.)


Supplementary Control Room

NAPS:

� Capability to�Reactor Shut-down

�Reactor Status Monitoring

� Incorporated in the design during advancedconstruction stage

� Common for both units


Supplementary Control Room…

Kaiga-1 & 2 and RAPP-3 & 4:

�Independent Power Supplies�Two approach routes from MCR to SCR�Provision of Local Control Panels (LCPs)�Relocation of LCPs in Kaiga-3,4/RAPP-5,6


Concluding Remarks

Design Safety Review of PHWRs, especially

of Narora NPP onwards has brought in

enhanced safety progressively into the

systems.

PHWR Safety Review Experience - Nuclear Regulatory Commission · 2012-11-19 · Atomic Energy...

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