1

S.C. Chetal

Director, Reactor Engineering Group

Indira Gandhi Centre for Atomic Research

Kalpakkam, India

PFBR Licensing Process and Experience

IAEA – GIF Workshop on

Operational and Safety Aspects of Sodium Cooled Fast Reactors

23-25 June 2010, Vienna

Contents

• Licensing process

• Deviations accepted with respect to safety criteria

for design of PFBR

• Civil Engineering Safety Committee review

• Project Design Safety Committee review

• Summary

2

Regulatory Consenting Process

Consenting Stages

Siting

Excavation Clearance

Construction Clearance for First Pour of

Concrete

Clearance for Erection of

Major Components

Commissioning

Operation

Decommissioning

Documents required to be Submitted/Reviewed

during Consenting Stage

Siting Site Evaluation Report

Construction • Safety Analysis Report (Preliminary)

• Quality Assurance during design and

construction

• Construction schedule

• Emergency preparedness plan

Commissioning • Commissioning Programme

• Technical specification for operation

• In-service inspection manual

• Fire hazard analysis

Operation • Safety Analysis Report (Final)

• Performance reports during commissioning

3

APPLICANTPROJECT DESIGN

SAFETY

COMMITTEE

CIVIL ENGG.

DESIGN SAFETY

COMMITTEE

INTERNAL SAFETY COMMITTEE

WORKING GROUPS

SPECIALIST GROUPS

ADVISORY COMMITTEE

FOR

PROJECT SAFETY REVIEW

AERB

AERB BOARD

CONSENT FOR

CONSTRUCTION

AERB

Scheme for Authorisation for Construction

PFBR Important Statutory Clearances

• Atomic Energy Regulatory Board

� Kalpakkam Site (October 2000)

� Manufacture of long delivery NSSS components (February 2002)

� Excavation clearance (July 2002)

� First of pour of concrete (December 2004)

� Permission for reconstruction of NICB raft (April 2005)

� Clearance for construction of super structure for NICB (December

2005)

� Construction of reactor vault and spent subassembly storage bay

(June 2006)

� Safety vessel erection (February 2008)

� Erection of major equipment (October 2009)

• Environment Clearance

� Public hearing by State Government (July 2001)

� Ministry of Environment & Forest (April 2003)

4

PFBR Technical Data

Electrical power 500 MWe

Primary system concept Pool

Primary & Secondary coolant Sodium

Fuel MOX

Primary & Secondary pumps 2

Primary Sodium Inlet/Outlet temp 397 / 547oC

Steam Conditions of TSV 16.7 MPa / 490oC

Shutdown systems 2 (9 CSR and 3 DSR)

Decay heat removal system 2 (4 x 8 MW – SGDHR

heat rejected from pool

to air)

Main Vessel Cooling From cold plenum

Steam Generator Integrated once through,

single wall, steam reheat,

4 SG / Loop

Fuel Debris Collection Core catcher

Containment Building RCC Rectangular

250 mbar

Deviations accepted with respect to

PFBR Safety Criteria for Design

5

Decontamination System

• Within containment,

use of water shall be

prohibited except for

biological shield

cooling.

• Decontamination

facility located inside

RCB from ease of

design. Usage of CO2

bubbling through

water for sodium

removal. H2 release

taken care of.

Sec. Na. piping

SGDHR piping

BSCS

piping

Fresh air Supply

Argon supply

Water system

Exhaust

Decont.

Facility

Containment Isolation Valves

• All lines penetrating the containment shall have at least one isolation valve

outside containment, except for those lines which are required to carry out

safety function in accident conditions.

• Exemption got for secondary Na piping valves (Ø 550 mm valve).

• Design of double envelope for RCB pressure

6

Core Disruptive Accident Classification

• “Core Disruptive Accident (CDA) should be considered as design

basis accident”. However, Project Design Safety Committee accepted

subsequently CDA as BDBE.

Nevertheless following design basis has been followed:

∗ Reactor Containment Building (RCB) is designed for pressure

resulting from expulsion of sodium burning in air of RCB.

∗ Site boundary dose meets the limits of 100 mSv allowed for

design basis accident (Category-4) on best estimate analysis.

∗ Main vessel is designed for CDA.

∗ Decay heat removal exchangers in reactor pool are designed for

CDA.

CDA + SSE

CDA + OBE

Not combined BDBE

� Issues which got resolved satisfactorily in short time:

• Concept of inter-connected buildings for nuclear island (Reactor

Containment Building, Steam Generator Building and Fuel Building) on

common raft.

• Lower elevation of turbine building (Design Basis Flood Level (DBFL) 100

years) in comparison to NICB (DBFL 1000 years).

• Rehabilitation of constructed raft consequent to sea-water

contamination (Tsunami).

• Acceptability of same DBFL consequent to Tsunami (Cyclonic condition

govern DBFL).

• Acceptability of concrete temperature for transient conditions of offsite

power failure.

Civil Engineering Safety Committee Review

7

Plant Layout

7.0M WIDE ROAD

TO MAIN GATE

R9.0M

36 36

NDDP COMPOUND W

ALL

EDGE OF GRADING

Location ofEffluent line

LA

UN

DR

Y

PFBR PLOT PLAN

8000

8000

34

35

5000

30M

455m

178M

132M

Sta

tio

nA

larm

Seco

nd

ary

SecurityDesk

BUS STAND

DOUBLE FENCING WITH CCTV & INTRUSION DETECTION

CCTV & INTRUSION DETETCTION

TO IGCAR

8m

28m

8m

5m

203m 318m

Compound wall ofNPC/DAE Facilities

8m 8m

5m

MAIN PLANT ZONE

OPERATING ISLAND

VITAL AREA

NO

RT

H

25. SITE ASSEMBLY SHOP

45. WATCH TOWER

44. LAUNDRY

43. CWMF CONTROL POST

37. CONDENSATE STORAGE TANK.

35. PARKING

36. SECURITY.

38. SEA WATER FILTERS

39. STORE.

40. OPEN SCRAP YARD

41. SPACE FOR SEWAGE LIFT STATION

42. TRANSFORMER AREA FOR CONVERTERS.

34. SEA WATER PUMP HOUSE.

28. ADMINISTRATIVE BUILDING

26. CONSTRUCTION POWER STATION

27. TRAINING CENTRE/ FUTURE EXTN.

29. CANTEEN

30. EMERGENCY TURBINE OIL TANK.

31. CONDENSER COOLING WATER OUTFALL.

33. SECURITY FENCE.

32. CONDENSER COOLING WATER INTAKE JETTY.

10. TURBINE BUILDING

24. CHLORINE BUILDING

16. HORTON SPHERES

13. LIQUID N2 STORAGE

14. DIESEL GENERATOR BUILDING-1

15. FUEL OIL STORAGE-1

17. STACK

18. SERVICE WATER PUMP HOUSE

19. FUEL OIL STORAGE-2

20. DIESEL GENERATOR BUILDING-2

21. RAW WATER AND FIRE WATER PUMP HOUSE

22. PACKAGE BOILER& FUEL OIL STORAGE TANK

23. DM. PLANT

8. ELECTRICAL BUILDING-2

9. SERVICE BUILDING

11. TRANSFORMER YARD

12. 220 KV INDOOR SWITCH YARD

7. CONTROL BUILDING

6. ELECTRICAL BUILDING-1

5. RAD WASTE BUILDING

4. FUEL BUILDING

3. STEAM GENERATOR BUILDING-2

2. STEAM GENERATOR BUILDING-1

1. REACTOR CONTAINMENT BUILDING

LEGEND:

� Issues which influenced design:

1. Operating basis earthquake PGA value

2. Combination of Category-4 DBE with SSE

• (Main vessel leak + SSE)

• Load combination in AERB code

• BDBE definition for PFBR- ≤ 10-6/Ry

3. Reactor vault design and construction aspects

Civil Engineering Safety Committee Review

8

Grade Level of Nuclear Island, Power Island and the level of Sea water

due to D.B.F. of 1000 / 100 Year Return Period and Tsunami level

occurred on 26 December 2004.

Finished Grade Level of

Power Island + 6.8 m

Finished Grade Level of

Nuclear Island + 9.3 mD.B.F.L. (1000 year return period) + 6.45m

TSUNAMI LEVEL + 4.71 m

D.B.F.L. (100 year return period) + 4.55 m

M.S.L.RL +/- 0.0 m

Project Design Safety Committee Review

� Codes Validation Sub-committee

• Reactor shielding

• Structural mechanics

• Thermal hydraulics

� Assisted by:

• 19 Specialist Groups constituted including one on erection of

reactor assembly component.

• Internal Safety Committee

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Project Design Safety Committee Review

� Issues which took long time to converge:

• Blanket monitoring

• Core support structure redundancy

• Neutron flux monitoring

• Acceptability of manual mode of reactor operation

• Decay heat removal system

• Reliability analysis of shut down system & decay heat removal

system

• SCRAM parameters

• Main vessel external pressure design margin

• Core disruptive accident analysis

• Containment design pressure

• Flow measurement in Blanket SA

• In-service Inspection of core support

shell to main vessel shell weld

• Drop time measurement of diverse

safety rod

• Experimental verification for confirming

of no lifting of subassembly in case of

SSE

• Safety vessel thermal insulation seismic

qualification

Reactor Assembly

10

Blanket Subassembly Flow Monitoring

Additional requirement :

• Flow through BSA will be checked in shut down state

using Eddy Current Flow Meter to minimise / avoid risk of

plugging during operation

Stipulations :

• Fresh Blanket SA should be checked for flow after loading

• Flow through selected 10 % of BSA to be checked during

every fuel handling campaign initially - Periodicity will be

reviewed after experience

Parameter Fuel Blanket

Peak Linear Power,

W/cm

450 350

Peak burnup, GWd/t 100 25

Clad hotspot limit oC

(normal operation)

700 680

Thermocouple 2 In few blanket SA

in first rowECFM

SV NITROGEN FEED

COVER GAS BLEED

COVER GAS FEED

P = 50 mb

RELIEF POT

P = 400 mb

P = 50 mb

RELIEF POT

BLEED

GASEOUS EFFLUENT

HEADER IN RCB

COVER GAS SPACE

REACTOR

INNER VESSEL

SAFETY VESSEL

MAIN VESSEL

REACTOR VAULT

INTERCONNECTION

P = 150 mb

RELIEF POT

TO RGEC

DEFENCE IN DEPTH :

MV-SV

SV NITROGEN

SAFETY VESSEL NITROGEN

MAIN VESSEL ARGON AND

ARE INTER CONNECTED

THROUGH A RELIEF POT

Limiting External Pressure of Main Vessel in case of an Event

11

Decay Heat Removal

• Decay heat removal system

reliability analysis.

• Independent design analysis.

• Effect of cyclonic wind

conditions.

• Experimental demonstration of

natural convection system.

• Incorporation of electro

magnetic pump in case,

tests/commissioning calls for

(Provision made)

Neutron Flux Monitoring

HTFC in Control Plug – Start-up &

Intermediate Power Range

HTFC in ICSA – Core loading & first

approach to Criticality

Fission Chambers - Power Range

12

SCRAM Parameters

SCRAM Parameter Threshold limitShutdown

system

Power 110% 1

Power/flow 1.1 (P/Q1) or (P/Q2) 1 & 2

Pump speed Nominal - 5% 1

Reactivity ±±±± 10 pcm 1

DND 1 & 2

Central subassembly outlet temp. Nominal + 10 K 2

Central subassembly temp. rise Nominal + 10 K 2

Mean core temp rise Nominal + 10 K 2

Deviation of individual SA sodium

outlet temp10 K 1 & 2

Reactor inlet temp Nominal + 10 K 1

• Issues debated a number of times in Safety Committee meetings.

• PFBR safety criteria comparison clause by clause with IAEA NS-R-1 “Safety

of Nuclear Power Plants: Design issued in 2000.

� [The publication is applicable for water cooled reactors. It is

recognized that in the case of other reactor types, some of the

requirements may not be applicable, or may need some judgement in

their interpretation].

� Specialist Groups appointed. Considerable work done.

� Exercise did not call for change in safety criteria or design.

� (Additional knowledge gained by IGCAR and AERB).

∗ Safety Criteria for Design of PFBR (Issued in 1990)

∗ Does it need modification?

13

Summary

• Safety review has helped to make PFBR design and

construction robust.

• Atomic Energy Regulatory Board has been convinced to give

stage-wise clearance based on track record and robustness of

approach and implementation.

• Better to accept conservative design recommendations of

safety committee instead of prolonging the issue.

• Safety Committee Members are competent persons who

have designed and reviewed thermal reactors. It has been not

easy to make them realise the different response for PFBR as

sometimes the system response and behaviour are quite

different from PHWR.

Thank you