SEISMIC MARGIN OF 500MWe PFBR BEYOND SAFE

SHUTDOWN EARTHQUAKE

S D Sajish Dr P Chellapandi

S C Chetal

Nuclear and Safety engineering GroupIndira Gandhi Centre for Atomic Research

Kalpakkam -India

Outline of the presentation

� Introduction

� Design basis ground motion parameters for Kalpakkam

� Input generation for seismic design

� Seismic design of critical components of PFBR

� Buckling analysis of PFBR reactor assembly

� Seismic qualification by shake table experiments

� Seismic capacity assessment of PFBR beyond SSE

� Ongoing/Future projects

� Conclusion

Introduction

GP

IHX

PSP

CSS

CSRDM

Primary pipe

InnerWall

OuterWall

SV

MV

IV

Roof slab

Base Raft

� Prototype Fast Breeder Reactor

(PFBR) 500 MWe sodium cooled

fast reactor in advanced stage of

construction in Kalpakkam, India

� All safety related systems and

components of PFBR designed to

withstand the extreme loading

condition in the event of an

earthquake.

� Seismic design of PFBR based on

the site dependent ground motion

parameters.

� Seismic design of PFBR follows a

dual earthquake design criteria i.e

design for OBE and SSE loading

Cross section of PFBR reactor assembly

Determination of seismic ground motion parameters � PFBR situated in a very low

seismic area

� No active fault within 5 km radiusof site

� Ground parameters determinedbased on deterministic method

� Values confirmed by performingprobabilistic seismic hazardanalysis

� Geological and seismologicalinvestigation of site

� Total 25 faults have beenconsidered

� Identification seismic potential offaults

� Mapping of historical earthquakerecords

Determination of Seismic ground motion parameters � Determination of the maximum earthquake potential of each f ault or source

� Site dependent attenuation formula for determining the pea k groundacceleration

� Determination of the design response spectra using statist ical methods

� PSHA to confirm the design basis ground motion parameters

� Development of uncorrelated design time histories in three orthogonaldirection.

Seismic design of PFBR

Direction SSE OBE

Horizontal 0.156 g 0.078 g

Vertical 0.104 g 0.052 g

FE model of PFBR NICB

Design time history (Horizontal) for SSE

FE model of PFBR NICB

Seismic design of critical components

� Safety critical components designed for both OBE and SSE bas ed ondetailed analysis and experiments

� Safety related systems include the following

1. Reactor assembly 2. shutdown systems

3. SGDHR system 4. OGDHR system

Seismic design of reactor assembly• Thermo mechanical properties at high operating temperatur e demands thin

walled structures for the reactor assembly• Large diameter concentric shells with Narrow gap filled wit h liquid sodium

• Response under seismic loading highly complex due fluid str ucture

interaction effect• Presence of sodium in the narrow annular gap generates large hydro

dynamic forces during seismic event• Possibility of buckling of thin shells

FE model Modal analysisSeismic analysis of RA

Seismic design of reactor assembly

Component Pm (MPa) Pm+Pb (MPa)Actual Allowable Margin Actual Allowable Margin

Main vessel 160 242 1.51 256 363 1.41

Inner vessel 155 230 7.6 164 345 7.0

Inner thermalBaffle

30 101 3.06 49 151 2.9

Outer thermalbaffle

33 242 1.56 52 363 2.2

Control plug 47 230 4.89 50 345 6.9

0.25 Hz 1.2 Hz 2.8 Hz 5.1 Hz 6.0 Hz 8.8 Hz

Shake table experiments of reactor assembly

� Displacement response using LVDTs and laser sensors

� Dynamic pressure distribution using pressure sensors

� Dynamic amplification using tri-axial accelerometers

� Sloshing response using laser sensors

� Strain measurements using strain gauges

Seismic buckling of reactor assembly� Large diameter coaxial thin vessels with diameter to thickn ess ratio ~500-800

� Seismic events impose high dynamic forces to these vessels e ven though the

static loading, under normal operating conditions is very l ow

� The presence of annulus liquid in the small gap between the co axial shells

contributes significant added mass to the adjacent shells

� Situation demands demonstration of buckling strength of un der seismic

condition

3D buckling analysis at critical instants during seismic excitations

Inner vessel Inner baffle Outer baffleMain vessel

Seismic buckling of reactor assembly

Component Buckling load factor

value Available margin

Main vessel 3.2 1.947

Inner vessel 1.9 1.46

Outer thermal baffle 3.2 2.46

Inner thermal baffle 3.0 2.3

Seismic analysis of core� To evaluate the stress distribution and the reactivity osci llation under seismic

loading

� The mechanical interactions among the subassemblies calls for complex

nonlinear analysis

� In house code CORESEIS used for the study of core seismic behavior

� The net reactivity oscillations due to horizontal and vertic al excitations are

found to be 0.11$ which is less than the specified limit of 0.5 $

� stresses developed the reactor core due found to be in the ord er of 100 Mpa

with allowable limit of 636 MPa

Figure.5 Seismic response of PFBR reactor core

Parts (N) Level D (N + SSE)

Pm Pm+Pb Pm Pm+Pb

Support shell 23 35 32.0 46.5

Stiffener 4.8 6.0 9.1 11.3

Top plate 8.2 11.2 18.3 24.3

Bottom plate 20.4 21.4 39.0 41.0

Seismic analysis of CSS

Collapse load estimation

Collapse load estimation of 1/5 th scale CSS model

For the normal load of 925 t, the load factor is 3.9 by experime nt, comparedto the theoretically predicted value of 2.4

Seismic analysis of primary pump

Schematic of primary sodium pump

� For the seismic analysis of pump, special purpose computer c ode PUMPSEIS has been

developed

� Transient time history analysis to determine stresses and t he possibility of pump

seizure during seismic excitation

� stresses in the pump shaft and pump shell are insignificant ( < 20MPa)

� The eccentricity of the pump shaft increased from 60 µ in norm al operating condition to

250 µ under SSE but < 400 µ

FE model of primary pump Mode shape of primary pump

Seismic analysis of CSRDM

� To determine the scram time

during a seismic event by

considering the drag and the

mechanical interaction between

mating parts

� To determine the stresses under

seismic loading

� Scram time under SSE is less than

0.62 s < 1 s

� Stresses in the components are

insignificant

� Results confirmed by full scale

testing in air

Scram time analysis of CSRDM

Mode:1 0.4Hz 1.8Hz

0 10 20 30 40 50 60 70 80 90 1000

5

10

15

20

25

30

Frequency (Hz)

Acc

elle

ratio

n (m

/s2 )

TargetAchieved

Seismic qualification of SGDHR

� SGDHR consists DHX, expansion tank, AHX, air

damper and piping system

� System is qualified based on the rules of class-

1 component

� Seismic design of all systems and components

are based on detailed analysis by response

spectrum method

� Piping systems are analyzed based on multi-

support excitation

Seismic qualification of SGDHR components

Mode Shape for DHX Model A ( 6.5 Hz)

(6.5 Hz) (20.1 Hz)

Dominant Mode shape for AHX (5.11 Hz)

Schematic Sketch of AHX (Type –A)

� Max Stress in DHX is 109 MPa while allowable stress intensity is 367 MPa

� Maximum stress in AHX during SSE is 140 MPa with allowable lim it of 362 MPa

Seismic qualification of piping systems

� Piping systems for SGDHR and OGDHR are high operating temper ature low

pressure thin walled piping systems

� High flexibility requirement due to high operating tempera ture ~ 5500C

� Special type of supports known snubbers/dampers are essent ial for the seismic

qualification of sodium piping systems.

� Realistic excitations are employed in the analysis by multi -support excitation

method and nonlinear time history analysis

ANC

ANC

VSH

SRT

SRT

SNU

SRT

SNU

SNU

VSH

VSH

ANC –AnchorVSH-Variable support hangerSNU-SnubberSRT-Rod support

(A)

(B)(B)

(B) (B)

(B)

(B)

(B)

(B)(C)

Seismic qualification by shake table experiments

� To demonstrate the structural integrity

and functional requirements of

electrical, electronic, instrumentation

and mechanical components

� Components are tested in energized

condition

� Capacity assessment by fragility tests

� Seismic qualification tests are based on

the guidelines of IEEE-344 and ASME

QME

� Tests are conducted using multi axial

shake table and electro dynamic slip

table

� Input excitations are generated as per

IEEE procedure

Seismic qualification by shake table experiments

control panels qualificationLighting system

Core seismic SV insulation panel1/10th model of MV with internalsHSB

Valves

� Tests are conducted for OBE (5 OBE ) and SSE (1 SSE) conditions

� Responses measured by accelerometers, strain gauges, LVDT , high speed camera etc

Seismic capacity of PFBR components beyond SSE

� Review of seismic design aspects of the PFBR based on Fukushi ma event

� Safety margin available in components and systems due to the design features

� Safety margins are assessed for all the safety related syste ms and components to

meet any demand beyond SSE.

� Most of the components except MV and SV have a safety margin mo re than 2

� Reactor assembly is the most critical system w.r.t the seism ic design

� Main vessel and safety vessel are most critical components i n the reactor assembly

� Plastic deformation and buckling are found be the predomina nt failure modes

Component Pm (MPa) Pm+Pb (MPa)Actual Allowable Margin Actual Allowable Margin

Main vessel 160 242 1.51 256 363 1.41Inner vessel 155 230 7.6 164 345 7.0InnerthermalBaffle

30 101 3.06 49 151 2.9

Outerthermalbaffle

33 242 1.56 52 363 2.2

Control plug 47 230 4.89 50 345 6.9

Table.1 seismic analysis of reactor assembly

Seismic capacity of PFBR components beyond SSE

Component Buckling load factor

value Available margin

Main vessel 3.2 1.947

Inner vessel 1.9 1.46

Outer thermal baffle 3.2 2.46

Inner thermal baffle 3.0 2.3

� Results indicate that a safety margin of 1.41 is available fo r main vessel against

plastic collapse

� Inner vessel is most vulnerable component for buckling with a safety margin of 1.46

� The safety margins are evaluated for a peak ground accelerat ion value 0.156 g

� Hence PFBR have safety margin of 1.41 for SSE

� So PFBR can withstand an earthquake beyond SSE up to a peak gro und

acceleration= 1.41 ××××0.156 =0.22 g without violating any safety limits

Table.2 Buckling analysis

Ongoing/future projects

� Seismic probabilistic safety assessment (SPSA) of PFBR to d etermine

the core damage frequency

� Development of seismic fragility curves for reactor assemb ly

components based theories of reliability

� Seismic reliability analyses of sodium piping and reactor a ssembly

components

� Scram reliability of shutdown systems under seismic loadin g

� Reliability based design for fast reactor components

Summary

� Seismic design aspects of safety related systems and compon ents of PFBR is

discussed with a focus on reactor assembly components

� PFBR is situated in a low seismic area with a peak ground accel eration value of

0.156 g

� The design basis ground motion parameters for the seismic de sign are evaluated by

deterministic method and confirmed by probabilistic seism ic hazard analysis

� Review of the seismic design of various safety related syste ms and components

indicate that margin is available to meet any demand due to an earthquake beyond

SSE.

� Reactor assembly vessels are the most critical components w .r.t seismic loading

� Minimum safety margin is 1.41 for plastic deformation and 1. 46 against buckling

� From the preliminary investigation we come to the conclusio n that PFBR can

withstand an earthquake up to 0.22 g without violating any sa fety limits

� Additional margin can be estimated by detailed fragility an alysis and seismic margin

assessment methods.