ML20216F990

Summary of 980225 Meeting W/Licensee to Discuss Issues Related to Review of Licensee IPE Submittal.List of Attendees Encl

ML20216F990
Person / Time
Site:	Crystal River
Issue date:	03/13/1998
From:	Wiens L NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9803190163
Download: ML20216F990 (19)

Text

,

i LICENSEE: Florida Power Corporation March 13,1998 l l

FACILITY: Crystal River Unit 3

SUMMARY

OF MEETING ON FEBRUARY 25, 1998. REGARDING INDIVIDUAL PLANT EXAMINATION (IPE) SUBMITTAL On February 25, 1998, a meeting was held between representatives of the i Nuclear Regulatory Commission (NRC) and Florida Power Corporation (FPC),

licensee for Crystal River. Unit 3 to discuss issues related to the review of the licensee's IPE submittal. .The licensee presented discussions addressing NRC questions'regarding the IPE and improvements which had been made to the  ;

Crystal River. probabalistic safety assessment (PSA) since the original IPE submittcl. The NRC representatives indicated that the meeting had been very useful in reaching resolution of questions concerning the IPE submittal. l Enclosure 1 is an attendance list for the meeting. Enclosure 2 includes j copies of the meeting handouts.  :

i

/s/

Leonard A. Wiens. Senior Project Manager Project Directorate II-3 i Division of Reactor Projects - I/II p Office of Nuclear Reactor Regulation Docket Nos. 50- Ws and 50-389

Enclosures:

1. Attendance List j

2. Handouts l cc: See next page DISTRIBUTION: E-Mail W rd Cooy SCollins/FHiraglia. TMartin docket File RZimmerman MTschiltz, RII PUBLIC JZwolinski LP11sco, RII i I

Crystal River r/f FHebdon OGC LWiens  ; I

-ACRS BClayton 0FFICE PDII-3/PM. s PDII-3/LA PDII-3/D s LWIENS h W b NAME BCLAYT0y*/ FHEBDONh DATE S/B/98 3 /4 /98 3 /13/98 0FFICIAL RECORD COPY DOCUMENT NAME: G:\ CRYSTAL \MEETSUM.IPE " ~ 9 m 3 n ~.-

9803190163 900313 M J u d; g1 d; @ gwI ys lll lll

• i e uIo 1 UNITED STATES

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20656-0001

%,,,,/ March 13,1998

> LICENSEE: Florida Power Corporation FACILITY: Crystal River, Unit 3

./

SUMMARY

OF MEETING ON FEBRUARY 25, 1998. REGARDING INDIVIDUAL PLANT EXAMINATION (IPE) SUBMITTAL On February 25,'1998, a meeting was held between representatives of the.

Nuclear Regulatory Commission (NRC) and Florida Power Corporation (FPC).

. licensee for Crystal River, Unit 3 to discuss issues related to the review of

-the licensee's IPE submittal. The licensee presented discussions addressing NRC questions regarding the IPE and improvements which had been made to the Crystal River probabalistic safety assessment-(PSA) since the original IPE submittal. The NRC representatives indicated that the meeting had been very useful in reaching resolution of questions concerning the IPE submittal.

Enclosure 1 is an attendance list for the meeting. Enclosure 2 includes copies of the meeting handouts.

Leonard A. Wiens, Senior Project Manager Project Directorate II-3 Division of React'or Projects - I/II Office of Nuclear Reactor Regulation Docket Nos. 50-335 and 50-389

Enclosures:

1. Attendance List

2. Handouts '

cc: See next page

Mr. Roy A. Anderson Ckt dTAL RIVER UNIT No. 3 Florida Power Corporation ec: . Chairman Mr. R. Alexander Glenn Board of County Commissioners Corporate Counsel . Citrus County Florida Power Corporation 110 North Apopka Avenue MAC-A5A Ivemess, Florida 34450-4245

' P.O. Box 14042 St. Petersburg, Florida 33733-4042 Mr. Robert E. Grazio, Director -

Nuclear Regula*ory Affairs (SA2A)

Mr. Charles G. Pardee, Director . Florida Power Corporation Nuclear Plant Operations (NA2C) Crystal River Energy Complex Florida Power Corporation 15760 W. Power Line Street Crystal River Energy Complex Crystal River, Florida 34428-6708 15760 W. Power Line Street Crystal River, Florida 34428-6708 Senior Resident inspector Crystal River Unit 3 Mr. Bruce J. Hickle, Director U.S. Nuclear Regulatory Commission Director, Restart (NA2C) 6745 N. Tallahasses Road Florida Power Corporation Crystal River, Florida 34428 Crystal River Energy Complex 15760 W. Power Line Street Mr. John P. Cowan Crystal River, Florida 34428-6708 Vice President, Nuclear Production (NA2E)

Mr. Robert B. Borsum .

Florida Power Corporation Frematome Technologies Inc. Crystal River Energy Complex 1700 Rockville Pike, Suite 525 15760 W. Power Line Street Rockville, Maryland 20852 Crystal River, Florida 34428-6708 i

Mr. Bill Passetti Mr. James S. Baumstark Office of Radiation Control Director, Quality Programs (SA2C)

Depadment of Health and Florida Power Corporation Rehabilitative Services . Crystal River Energy Complex 1317 Winewood Blvd. 15760 W. Power Line Street Tallahassee, Florida 32399-0700 Crystal River, Florida 34428-6708 Attomey General Regional Administrator, Region 11 Department of Legal Affairs U.S. Nuclear Regulatory Commission The Capitol 61 Forsyth Street, SW., S'#te 23T85 Tallahassee, Florida 32304 - Atlanta, GA 30303-3415 Mr. Joe Myers, Director Mr. Kerry Landis Division of Emergency Preparedness U.S. Nuclear Regulatory Commission Department of Community Affairs 61 Forsyth Street, SW., Suite 23T85 2740 Centerview Drive Atlanta, GA 30303-3415 Tallahassee, Florida 32399-2100

]

i MEETING ATTENDEES i i

CRYSTAL RIVER INDIVIDUAL PLANT EXAMINATION MEETING FEBRUARY 25. 1998 l l

ME USSMIZATION i Len Wiens NRC/NRR/PDII-3 Jim Baumstark FPC/0uality Programs

]

Sherry Bernhoft FPC/ Licensing  ;

l

! Mark Averett FPC/ Safety Analysis )

l Mike Rencheck FPC/ Director Engineering i

Jack Tunstill FPC/ Licensing John C. Lane NRC/RES/PRAB i

Mary Drouin NRC/RES/PRAB l i l l l

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Crystal River 3 PSA/IPE g.

m-a; . . .

"i.g$ Mark Averett aw "maz,gn::s e

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• i. Flodda Power Corporation v,;

! Outline m- Jd m@m~4t,7d m

p* y a Background / History j

" 7 d a Resolution ofNRC Comments on IPE l mm. a j

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.. %, a Use of PSA at CR-3 t

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.s m Conclusion 3 5 7 ,s, u Q j e4 ggr  :

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Objectives

,[yg a *j a regardingDiscuss the IPE any past questions / concerns I 1 d a Discuss improvements to the CR-3 PSA

4 . . .

at; m>0 q since the original IPE submittal

• i a m Ji a Bring the CR-3 IPE to closure with respect

"*ym.

• i to Generic Letter 88-20 i

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Background / History En da g o. nw ddry 6/84 718 7 11/87 11/88 1/89 g i%.* ' .]i:3 _

Gener6c NUREGICR4245 CR-3 PSA NRC Review of I ' ;il CR-3 PSA CR-3 PSA ., 2 Letter a E " Submitted to j Begun Published .,

NRC 88-20 CR 3 PSA E1 hj Published Published E ll- / tJJ E;_ sj E[ i EJ' j:]

jij. 363 9/95 11/95 4/97 7S7. 2/9B

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3 : y I

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. CR-31PE FPC ~N" d

g p+,iM <w Submitted Rec ed  :% Response CR IPE to NRC E ~g w to NRC from NRC

~E~- Received Letter l

1 1

l CR-3 IPE vs. PSA p"g a CR-3 IPE was a snapshot look at risk for air.; F. CR-3 as it was configured in 1993.

![ $ m CR-3 IPE core damage frequency =

Eb i 1.4E-05 per year.

"p arm] s CR-3 PSA is a living PSA representing the l} g risk for CR-3 as the plant is currently configured.

1 I

l NRC Comments on CR-3 IPE

Pt4 from 4/97 Letter m ze er m mus a Front-end s* %

": 4 j u Human Reliability Analysis

, 4 at J u Back-end awa "i 1 m ~ General e  :-

he?3 . . . .

l

NRC Front-ead Comments  ;

1 l

l

? f n Initiating events

$/09 m Failure data y' y a] a Internal Flooding

,a i g a Common-cause failure:,

. g y!

% 14 8;;$:

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Resolution of NRC Initiating gf Event Comments h.?l m

f n Addition ofloss of DC power u[! ' i;j events (CDF increase of <4E-8 per year)

I- : m Loss of NNI included in Loss of MFW

, a "j, j u LOCA frequencies aw a hY.1"l 5 5 gr 7;

Resolution ofNRC initiating  ;

Event Comments (cont.)

1 th' g;g -f $;h D&W PSA LOCA Frequency Companson g'f p L,OCA

. - + w -C, R 3 s

. B&W Urut .-

-Bat" Unit -

.-2'

- B&W Unit -

3

B&W Unit ~

9 4 /'

)

E S ' $1 SrnalLBreak LOCA 2.0E43 5 0E-03 1.4E43 2.3E 03 3.6ET3~ '

g Medium-Break tOCA 50E44 3.4E 44 lh 1.0E 03 3.6E-04~ 3 0E 04 E 22 < .;ih! Large-Break LOCA* 5.0E-05 1.0E-04 3 4E-04 1.4E-03 1.0F.-04 ,

E .- , .jeg l 5 " '"

I j

• sensitivity case for LBLOCA frequency of 3 OE-04 per year resulted in j the core darnage frequency increasing from 1.39E-OS per year to j 1.45E-05 per year. '

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s' ' Resolution of Failure Data n.

um  :

Comments EI(Y m? wi ki-a ISLOCA check valve rupture failure rate m; jy aL q[l4 + originally assumed 10% of check valve ruptures occur  ;

"; ig internally which translated to a failure rate of 1.15E-7 I m; - 3 per hour ae e a(

+ Aggregated industry data gives a check valve internal l

",pU:'d gdi rupture rate of 7.55E-8 per hour a e7 m Turbine-driven EFW pump failure to run was increased to 1.27E-3 per hour

+ increased IPE CDF from 1.39E-5 per year to 1.45E-5 per year

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I Resolution ofInternal Flooding '

L Analysis Comments  ;

g a 1 m i j h Y {@found m Plant walkdowns little potential for accumulation for the intemal of water in flooding anai

[ '4 ' " the rooms due to dmin problems because of the 7$ih open design.

abd@f a Maintenance-induced floods were not modeled. ,

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Generic industry internal flood data gave a maintenance-induced internal flood frequency of i

N 2.9E-3 per year. Applying this frequency to the  ;

existing internal flood analysis resulted in an increase in IPE CDF of <3E-7 per year.

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.ac , Resolution of Common Cause i

, Failure Analysis Comments u

u3 %

my a Common cause failure analysis completely me, n r.

il redone aj "l  ; + MGL approach ( as described in NUREG-4780 l j and NUIEG-5801) l sj + Common cause failure rates changed littie from the original IPE values f;jpg a 7 m Coromon cause failure events for the EFW pumps was added to the CR-3 PSA model

(<lE-8 per year merease m IPE CDF)

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i Resolution of Common Cause BcJ Failure Analysis Comments (cont.) {

x 4h s Comparison of Common Cause Failure Probabilities g]ry/:50;@3 g& j

'/ CR-J IPE and CR-3 PS A ffk,f[d N l

g. 'f Component Failure Mode - IPE Beta Factor PSA Beta Factor i g] ,

Desel Generators Desel Generators Fad to start Failto run

.021

.021 010

.016

, Decay Heat Pumps _ F1ll to start .046 .018 g 3,5 Decay Heat Purnps Fail to run 046 012 g g l!J. j;[/jServrce Water Pumps Fait to start .012 .016 g>. 4 :1 Service Water Pumps Fail to run 012 017 l g Makeup Pumps Fad to start 071 .054

(( Makeup Pumps ggyg!gl- MOVs Fad to run Fail to open 071

.033 018

.031 Check valves Fail to open 025 027 Eg 1 1

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4 Human Reliability Analysis s.i m

Comments g$y m ' %fj u Non-proceduralized recovery actions F

m ,

4 m Performance shaping factors

= r M. -1 m Time available for operator action m: s 5" sj m: 4 E:

3, y :f E l .m d E 9d Q ' d m:g agp$

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Resolution of Human Reliability Analysis Comments )

m; c: m Only one non-proceduralized recovery

=4 E:

%!2 action in the CR-3 IPE "W

Eu ,

]a a Human Reliability Analysis completely Er ms redone Eh $

'idi + Combination of SHARP 1 analysis for cognitive sS5E?? portion of operator action and Swain-Guttman i

a j (NUREG/CR-1278) for the execution portion of the operator action ia + Used in the Davis-Besse IPE

.oj y:

s.

Resolution of Human Reliability kEas1 Anal Ysis Comments (cont.)

gi &bMM E k b 20 Comparison of Operator Error Probabilities ED? ' W CR 3 IPE and CR 3 PSA E 11 4 ' *j Ei.. O Operator Action IPE Probability PSA Probability EO6' 3 Operating crew faias to make transition 1.0E 03 2.3E-04 E[' q d to high pressure recirculation El Operating rvew fails to rnake transition 1.87E-03 9.1 E-03 E( +-'^ >glj1to low pressure recirculation E G ^ r ' s@ Operating crew falls to make early 5 BE-02 9.1 E-03 E \i i bi transition to high pressure recirculation E kpsiirhil' _or to isolate EANST E EN3' yi Operating crew faits to switch cooling 1 OE-03 8.8E 04 Ej h'l sources to the makeup pumps E Operating crew fails to manualty 1.0E 03 61E 04 isolate makeup pump recirculation line

/

Resolution of Human Reliability l Analysis Comments (cont.)

wei u There are 8 different failure mechanisms in m;p (g '

the cognitive portion of the HRA which 7M allow for a wide variety of PSFs to be taken d?s into account.

[jav;jj u Extensive documentation of each operator "N action. Times based on thermal-hydraulic i calculations, discussions with operators and training personnel, and simulator runs.

1 l

l IPE Back-End Comments a 4 m;4 !M m; fd a Sensitivity study

!{

m' k] a Source term 1 .

=;; q m Contamment seals ashud! u Containment isolation details an geg::=l m Containment Performance Issue of

" Hydrogen Pocketing and Detonation ini ,

1

/

Containment Phenomenological Event I@ 9 m

e yev ,y ap 4

$1 )!

Tree (CPET) Sensitivity Study m CPET Events examined

+ CHRavailability

+ Hydrogen burn probability

'$ ' f + Induced RCS rupture probability

"[q . + DCH probability lk' M + Coolable debris bed probability 5 m Overall profile of CPET end states relatively unchanged

, Source Term Estimate for Station av ,4:: Blackout E@a'9' m Plant-specific severe accident progression sjf

":y lj model and analysis (STCP/CONTAIN) ,

I ; a Compared to NUREG/CR-4551 analysis for EU# O Zion - CR-3 iodine source term lower, my d ae  ; cesium higher, well within uncertainty l 5%w$j an t il bounds l Ei j u

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l Containment Seals gyggj u Seal performance as a function of fgjj temperature was compared to the

= ); ) +emperature profile of the severe accidents )

1 as modeled in the Level 2 analyses. i

,j ,i h a Seals were found to survive until late i gWij the most severe accident (SBO) such that i e their failure would make little change in the Level 2 results.

1 Containment Isolation

=:r .

ER _S my

'{g j} muEach Provided containmentmore isolation details path was in 2/12/98 letter I n, modeled and failure probabilities calculated Ed m e , 4M under different conditions (e.g., SBO, loss m; v#

mu o f m.strument air) u m: ' s Containment isolation cut sets were E o appended to the core damage cutsets s

Containment Performance Issue l l,ihJ,! u Hydrogen Pocketing and Detonation

! , 3 m NUREG/CR-5275 - evaluation method  !

k:) m NUREG/CR-4803 - application of method g.

i; rg? -g.g for Bellefonte, one problem compartment jj a CR-3 containment design very open - no

'"jsi compartment comparable to the Bellefonte compartment found 1

i 5 .

General Comments e' -1 3 '

- u Plant imprncments and insights l ap"-@h

", ' [

nI ncorporatica of PSA knowledge into plant  !

m .

er , :1 '

operatwns m;. m "l , j u Vulnerability definition m> d g .h . --, TJ EhhhfI

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l lgj., Plant Improvements and Insights m Improvements

"[ ~ y + Flush Water redesign "i 9 + BWST refill procedure m> a m! ai a Insights n

l gj av[j + Feed-and-bleed capability

$ + Diverse cooling systems I + Byron-Jackson N-9000 RCP seals

$a@Y gr y Incorporation of PSA knowledge hqis into plant operations an/ r #y a t$ u a[t Risk Monitor - on-line and outage

"( W3 m On-line system outage schedule risk Ei j assessment m >

"; , j u Maintenance Rule E$'j u Training synm l %. a CR-3 PSA Summary Document

,. s a PSA Web site, posters, cards

. 1

. . i i

Vulnerability Definition

% m The search for vulnerabilities is more of a process NE! than a threshold. Review of the core damage  ;

[:$ cutsets looking for sequences with unusually high I "I ] frequencies, sequences which reveal some

' }g ,' heretofore unknown dependency, and risk-Ena uf significant sequences which can easily be reduced to risk insignificance via a procedure change or l 5

minor hardware change consisted of FPC's review B@y of the IPE results for vulnerabilities.

l 1

e L Use of PSA at CR-3 ww i a.wm,

=BM! u Plant risk awareness - risk mo. i':ar, training,

=E aw a ed mamtenance rule, publications wq

" Ol; a Risk perspectives for licensing, engineering, ll av ~3 9 operations, maintenance, scheduling, and se 1 a, . .;;;

training atuvities "fy a Management support for active use of PSA 4 in decision-making f

s

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Conclusion l

{2d u Intent of Generic Letter 88-20 satisfied

$h8 m FPC management plans for PSA to continue f yll to grow as an integral part of Nuclear

, Li Operations activities

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Questions E

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