ML19325E884

Provides Initial Response to Generic Ltr 88-20, Individual Plant Exam for Severe Accident Vulnerabilities. Formal PRA Program Began in 1987 W/Final Intended Product Being Full Scope,Level III,plant-specific PRA

ML19325E884
Person / Time
Site:	Maine Yankee
Issue date:	10/31/1989
From:	Frizzle C Maine Yankee
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CDF-89-143, GL-88-20, MN-89-126, NUDOCS 8911090310
Download: ML19325E884 (16)

Text

f; v

)'

,e M.m MaineYankee E@RITrNRif7iiEGIig(}lgggg

,t

{f EDISON DRIVE e AUGUSTA, MA!NE 04330 * (207) 0224668 October 31, 1989 HN-89-126 CDP-89-143 UNITED STATES NUCLEAR REGULATORY CO!9tISSION Attention Document Control Desk Washington, D. C.

20555 Referencesi (a) License No. DPR-36 (Docket No. 50-309)

Subject:

Initial IPE P neric Letter 88-20) Response Gentlemen:

This letter provides Haine Yankee's initial response to Generic Letter 88-20, Individual Plent Examination for Severe Accident Vulnerabilities -

10 CFR 50.54(f).

Maine Yankee began a fornal Probabilistic Risk Assessment (FRA) program early in 1987 with the final intended product being a full scope, Level III, plant specific PRA.

The program involves a " phased" approach which will yield several useful intetuediate products. One of these products will be the Individual Plant Examination (IPE) as described in the Generic Letter 88-20.

The Maine Yankee PRA program, however, recognizes the benefits of additler analysis in the severe accident area and, as such, extends beyond the requirements as defined by the NRC at this time.

Phase I, which was completed in April of 1989, represents a Level I internal events PRA. Phase II, which is in progress with a late 1991 scheduled completica, will extend Phase I to both a Level II and Level III internal events analysis. Phase III, scheduled for completion in 1994. Will ine1Lde analysis of external events as covered by Levels I, II, and III.

The Maine Yankee PRA is being developed "in-house" by Yankee Atomic Electric Company engineers with reviews being conducted by both Yankee Atomic and Haine Yankee Atomic Power Company.

tlOOf A9?'090310 891031 PDR tn9CK 05000309 i

P PDC l~

fMM4PJ

b I

MaincYankee i

L p'

f United States Nuclear Regulatory Commission Page two i'

Attentioni Document Control Desk HN-89-126 i

i Existing documentation of the Phasa I PPJ. effort includes a single volume surnary report and more extensive calculation notebooks.

Attachment A provides the Table of Contents of our Phase I report.

Phase II documentation will be similar as it will largely extend the Phase I results.

The completion l

of Phase II of our PRA development effort will generally provide the information requested in Generic Letter 88-20 our summary report and calculation files / notebooks provide the "two-tier approach" to UE documentation discussed in the generic letter. The content of this documentation meets the intent of Generic Letter 88-20 and NURUG 1535 reporting guidelines.

In order to levelize our use of internal resources, and since it will be most efficient to perform all external event related analyses together, our present Phase II work plan does not include interna.' flooding ualysis. We plan to include an analysis of potentisl internal f1)oding events with our Phase III work which is scheduled to begin in 1992.

Haine Yankee believes that the maximum benefit of a PRA (or other IPE type t

analysis ) lies in the insights gained and in appropriately applying that i

is information to the design and operation cf the plant. While our Phase I analysis did not identify any significant risk "out-liers" or

" vulnerabilities", it did, of course, identify certain key sequences representing the top contributors to core damage risk.

Maine Yankee is currently pursuing possible options to reduce the likelihood of occurrence of these events. Various means of integrating PRA use into daily plant operations have been and are being implemented. We intend to continue this effort in conjunction with Phase II.

l In response to Generic Letter 88 20, Ha;.19 Yan).ee 11 submit a plant specific Level II internal events PRA by December 31, 1991. We propose to

[

submit the requested internal flooding analysis on a later schedule to be coordinated with Phase III of our program.

Our December, 1991 submittal will I

4 -

include a single volume Phase II summary report as well as a discussion of PRA application at Haine Yankee. More detailed calculation files and notebookr supporting our PRA efforts will be maintained by Maine Yankee / Yankee Atomic Electric Company and available for NRC inspection or audit.

We believe a meeting with the NRC staff to discuss our ppa. program further would be to our mutual benefit. We will be contacting the NRC Pro]ect Manager for Haine Yankee in the near future to arrange such a meeting.

4 g.

06511-RPJ

o7

<t

~r MaincYankee 3

1, 3

'I

.c i-United States Nuclear Regulatory Commission Page three

. Attention Document Control Desk HN-89 126 c

I t

We trust the precedir.g provides the inforr.ation you re4 Jested in Generic l

l' Letter 88-20 regarding Haine Yankee's methed, approach 4nd schedule for performing an IPi!.

Very truly yours, j

7 i

6 HATWE YANK ln Ab4 Z/

Charles D. Frizzle i

President i

GDW SJJ l

Enclosure I

c: Mr. Richard H. Wessman Mr. William T. Russell I

Hr. Eric J. Leedu Hr. Cornelius F. Helden i

Hr. Clough Toppan j

STATE OF HAINE i

i Then personally appeared before me, Chteles 0. Frizzle, who being duly i

sworn did state that he is President of Haine Yankee Atomic Ps.ter Company, I

that he is' duly authorized to exocute and file the foregoing response in tt.e name and on behalf of Haine Yankee Atomic Power Company, and that the statemen.,s therein are true to the best' of his knowledge ar.d belief.

l L

/

/

/

.i

/ %s.bJ C i.

tNI th j

u Notary Public

[

t UTj a conus,UA F CAsioNauAyilm y y,,,, ;

.g l

L, 1

l l

l i

1-i le I

tr.

. s

_f, ff Y

N $lQ A W. -

y wt9-L, s* N.

u.,. e

,ty)9 g,[d, 3 -

a,

.e ll; e.?Y 7/;;.6 3., =W y

m-.9 u >. n g't y

..n p

r

+1

. '..e

, 'j l

';s

-1 c J

c; -

- N p.

c' p Oy W

, ; t; v

., m,

i s"

y

.e e

s j.p.

6 d@'@$s%uqsd F 7 E.. N N ' (* L.

WaineYai#&

? :k

' r/,

' f. E, y

7 7

6 - ; e,;.g 4

e q

-1. 9

- p,

. i +

, l.. -,

g {

,4 z e-s-

i f,

'1 i '

O,u J L,,,

y

- ' i i

g i

y p. a f.",p ' m

• -n',

As e

.. f -

s c

/,, (1,

.)a 4

i K v. A L ',.. M T

'..r 4..

t,7 s

'g'.,

i z,i Q vih, v'

.y-

,.1 s

j=

8

. q )n, s

,m,...

g' e

~

8 G.

.,,4,3 i

,L

,1

s-W -

'r.,

UJi

+ t r

, [g u.y

.g*,

i

,s

\\

,, y, f.>. } ',

e.

^

,!-)V, C h.*..

m f( ' ? '.

t s

r L.

i 4

1

(

7 1

e

' " 3,$

fl' l U '

6 14].-

4,. eipt,,'

, a.,g.'

'f*

'\\'

.a,.

i ii s

q_

i;4 s.

'r ' i.

q.W e.,s, m-g. 3

3. i.s i

,, Yp f d, Q,'y& Lw &e,, - [< v,.c,l Cf r

. ' ' ;e' i

t

,n p sy v s

' f,I' llpf,[ @yfj $'6 ' i

/

a?

[/ h 1,[

- q[*

,{

[*'

j

}p,yW 4, y *{- i ho[1.

6. +,'.,F "

i i

5 o

..t ga

-,4 Ei-e

. 3

,Y<

{

% 9 4, J . t ! '., \\ ',.

r S

Thi

.Ir 9 l%

h,f. q e.4 3h,j o.

[

Y..h b b ~.

.~ '

b

a. g -

-F

,c-

,.c,

. t 6 6+

/

1 6,

yW 4ty.,[wy,

t i t

i 3

y g

s ok d_ E

,1

,.-s 3

d..r w;:

- ([go.[f i h. =,<

t

.t (, a

.4.,. f,p 's "% A.;j.

i i l,Q/, A.x.g4' x

t 7l-

"% ; o O,. b.. g,Si. ;.'t iip,g,

,, c.

'),.

. ATTACM3ff A-

, o w-p.,

i

,?(<

s m.; -

a.t

.,'l..*,.i.

' i, e

s r'

s

( $ l,', 'i h., j @,A -

'd

'I "t

'f 4We ) J't,j, a /il, t

.~s,

?i

- s p '(3 -, v. g f

gut- & ;a<

4

,, 6/

a._

r

+

v 4

(

'J' 1

t b,[i,)Q.. MM Q:

3 g+ j 4; y. a, % r,

)[.3,,1 3

..n 5

.+

y r-s r

.,t l ' dp m w} >3, i s,, q

.i $ /4 k -

r i. ' '

E f', i.-,y,,.-

-'1' 1

,r

[,{ a-3 {.

/

g

'A s s i.3-J '! !,.,

c.

i

$ $,j m,d % s jQ

-,l ?

t

'y S a

y q q.q,.

V j'.; (;

,. M..O...q N,,..,' st',

r b,, ' A 3 p. -

y y.

s t.

.)

s 1

d.

.--)<

n 9, s,

'n c.,,

q., 4.

i s.

3 7.,

s.

v,.

w..,

.s>

, )n(?.;M...p

.. p ( L'c i

1 e.. g' g,.,

}>

.- ( f ' L g J J,a.,, - "

u.. m.

p i

\\',? r>,,

J HAIllt YABEE PROSABILISTIC: RISK ASPESSHDft v

e,

+,.

e.- :

.co,

V'

' INS I.

r 7.

E

',y

,, f.., U,-

J

)

.. =

p; lb9....:

d

n,{' ynw1 i t

a m,

p_..

o r

9aJ. 47 v.

. :._ r e

^ >

- gyy ' g

%'.k.,

-g ; -

--p

,.e y,.

.' o a.g 3.,

m t','

l Wp/ J.1 g --

• ; ' y,. -

i p --

a-

"^

'h

[.r.piy; P.,.

g

' 7

j.

t i.

s 4 e.n.t,.

4p.7 s

y r..

.g

. lhl R e,d,,,1, m

. l h, u. %- g s

- '. ;" ' g/,'s v.

's' f

A E?_,y4

.v.'.

qQ:,

y'

,A

., u

.f ll > <g'.; ' Q. < 9 ;,

3

},

.3

. jy < m< :

o e

s

- re, n s

f k py:

.1 s

, 'e

+

e t-g g 3, r

r

%, ;krc,,n u

.0

,.fe

-1

' ((

4,,g s )

+

f'. - Iq :_,

1 l

4

....1v. 4.,i. J

"' h 'p m,,p,f'

,;M

• +

s i

c Q

s.M a&\\

,*4 },., q,,

J.'

1 {, _. -

ai ie

-l,

,,i.,'l_

],

.f r,. _*

.),V 0 7 l

.-h,,- '13s

. \\., %

,,1 5

_s

,., - ' op g

.o

.7..- ',P.9,,/ ' r$.

s..u.'

1 p

y

'(

z h

_g' af m

.o.

4,.

t {.

- h g' L

. p' h

t

~

' j'

'I l.'

,5 i

h

^ '

U,a y.pe N N'y,',

'p

. y2y a.

- th y f

m m

- a n ;,,,,;!,

n,,._

w, _..,

.a w

l YABC-it*6$

April 1967 w

W i

c.

i i

MAINE YANKEE FRA l

>i

- FEASE I REPORT -

n i

Prepared for Maine Yanket. Atomic Power Company 7

f i

Prepared By:

4 M P

D. C. Fan V

(Dat's )

i M. k

/!/!8/

Prepared By:

S. M. Follen (Date) l WINf}%ff/1/

h/ 7/01

)

Preparut 5yt K. Ghahrame.nl (Date) i i

e

[/2,!N J

v Prepared By:id. 4. O'Btien (Date) r t

hw 87 Prepared By:

We K. E. St. John /

(Date) l l

7!f9 Approved By: h R. I:hapman, 4 nager A

er

1. (6ste) j-Safety Assessment Group Y

8h

[

I Approved By:

I. C. STi er, Director

• (Da(e) '

f j

l Nuclear Engineering Department l

3' Yankee Atomic Electric Cotopany Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740 j

.6606R l

l 1

ARETRACT b the ' extent that this report describes Phase I of a multiphase FAA development for Maine Yankee, it is an intermediate report. To the entent that Phase I is a complete, reasonably detailed, limited scope level 1 Plus Probabilistic Risk Assessment, this report sume rises a !Laal and useful l

product.

j l

l Phase I results demoastrate that, in terms of calculated core damage l

frequency induced by " internal" initiating events from normal at-power operation, Maine Yankee is not e " risk outlier." the analysis did identify l

certain key contributors to calculated core damage frequency as well as

{

related uncertainties which should be pursued in latter phases of this PRA l

development.

i i

i terhaps the most useful Phase I result is a capability and an improved f

understanding of risk. Phase I produced workable, useful, and readily enpandable risk models as wel. as a strong "in-house" PRA capability for Maine Yankee.

It provides considerable in6'.shts to risks at Maine Yankee.

I t

I a

i i

r 0

e i

-111-6606R 9

e m

E I

l

.\\

i TABLE OF CONTENTS t

i DISCLAIMER 0F RESPONS!b!LITY......................................

11 AS S TRA CT..........................................................

iii i

i LIST OF EARLES....................................................

vii l

t i

LIST OF FIGURES..................,.................................

xii I

A CKN 0WLEDOM ENTS......,............................................ xiv 1.0 INTRODUCT10N......................................................

1-1 1.1 Background..................................................

1-1 1.2 Report Furpose..............................................

1-1 l

1.3 Report Organisation.........................................

1-2 l

2.0 SUM 4ARY AND CONCLUSIONS...........................................

2-1 l

i 2.1 Score.......................................................

2-1 2-2 2.2 Core Damage Frequency.........................-

2.3 Key Uncertainties..................................

2-5

~

2.4. Plant Damage State Frequencies..............................

2-7 2.5 Conclusions................................................

2-8 3.0 SC0FE.............................................................

3-1 4.0 AFFROACH AND METH0DS..............................................

4-1 ii 4.1 0v e rv i e w....................................................

4-1 l

4.2 Critical Safety Function....................................

4-8 4.3 Initiating Events...........................................

4-9 4.4 Suppcrt Systes Event Trees..................................

4-19

?

4.5 Frontline System Event Trees................................

4-22 4.6 P l an t Dama g e S t a t e s......................................... 4-24 4.7 Event Sequence Success Criteria.............................

4-26 4.8 Systes/ Function Analysis...................

................ 4-27 4.9 Data Analysis...............................................

4-36 4.10 Euman Actions...............................................

4-48 4.11 Sequence Quantification.....................................

4-52 4.12 Sensitivity and Uncertainty Analyses........................

4-56 I

?

-iv-6606R i

d 6

t 4

TARIA 01 camMrs (Continued) i Zass 5.0 CRITICAL SAFFTY PUNCT10NS.........................................

5-1 i

5.1 Definition and Use of CSFs..................................

5-1 5.2 Development of CSFs.........................................

5-1 i

5.3 E0Ps/FRGs...................................................

5-2 l

5.4 CSFs Used in Evaluation.....................................

5-5 I

5.5 Reactivity Contro1..........................................

5-5 l

5.6 Reactor Coolant Sys tem Inventory contro1.................... 5-6 5.7 Core Cooling................................................

5-6

)

5.8 Containment Integrity.......................................

5-7 6.0 IN I T IAT I NG EV ENTS................................................. 6-1 6.1 Summary of Maine Yankee Ope rating His tory...................

6-2 l

6.2 Loss of Primary Coolant.....................................

6-4 6.3 General Transients..........................................

6-19 6.4 S up po r t Sy s t em De g r a d a t i on..................................

6-32 6.5 Loss of Secondary Coo 1 ant...................................

6-83 l

t 7.0 EV ENT S EQU ENC E S AND TREES.........................................

7-1 7.1 Support Systems.............................................

7-1 l

7.2 Frontline Systems...........................................

7-3 8.0 SYSTEM M0DELS.....................................................

8-1 t

t 8.1 Secondary Fast Remova1......................................

8-3 l

8.1.1 Condensate..........................................

8-4 d.1.2 Main Feedwater......................................

8-9 8.1.3 Emergency and Auxiliary Feedwate r................... 8-14 8.1.4 Secondary Steam Relief..............................

8-21

{

8.1.5 Secondary Isolation.................................

8-28 l

l 8.2 Emergency Core Cooling Systems..............................

8-33 8.0.1 Eigh Pressure Safety Injection......................

8-34

8. /. 2 Low Pressure Safety Injection.......................

8-40 8.1.3 Safety Injection Tanks..............................

8-45 h

8.1.4 containment spray...................................

8-48 l.

8.2.5 ECC8 Signals........................................

8-54 t

8.3 Electric Power Dastribution.................................

8-58 i

-v-6606R i

TABLE OF OtaffMfTS (Continued) lass i

1 8.4 Cooling Water...............................................

8-68 4.4.1 Component Cooling Water.............................

8-68 j

8.4.2 Se rv i c e Wa t e r....................................... 8-75 t

8.5 Compressed Air..............................................

8-78

)

i 8.5.1 Instrument Air......................................

8-78 8.5.2 ErcV Air............................................

8-42

)

i

\\

8.6 Other Systems...............................................

8-85 8.6.1 PORVs/RCS Pressure Relief...........................

8-85 l

8.6.2 Emergency Boration..................................

8-89 8.6.3 Reactor Protection / Reactor Trip.....................

8-91 8.6.4 RVAC................................................

8-92 1

8.7 Containment Isolation......................................

8-96

}

8.7.1 Containment Isolation Va1ves.......................

8-96 8.7.2 Containment Isolation Signa 1.......................

8-101 l

?

9.0 DATA..............................................................

9-1 i

i 9.1 Introduction................................................

9-1 l

9.2 Ov e ra l l De s c r i p t i on.........................................

9-2 l

9.3 Component Failure Rates.....................................

9-6 9.4 component Maintenance Data..................................

9-10 9.5 Component common Cause Ta11ure..............................

9-14 9.6 Comparison of Generic and M61ne Yankee-Specific Data........

9-16

't 10.0 RUMAN ACT!0NS.....................................................

10-1 j

l l

l 10.1 La t en t H uman A c t i on s........................................

10-1 l

10.2 Dynamic Numan Actions.......................................

10-2 i

10.3 Sequence Recovery...........................................

10-3 11.0 PLANT DAMAGE CLASSES..............................................

11-1 12.0 PLANT MMAGE STATE QUANTIFICATION.................................

12-1 l

12.1 Quantification Methodology..................................

12-1 12.2 Quantification Results......................................

12-6

13.0 REFERENCES

13-1 9

APPENDICES A Interfacing Systems L0CA.................................

A-1 B Event Sequence Diagrama..................................

B-1

-vi-1 i

6606R 1

I L

- - - ~

i

• ~

LIST OF TABLES thabar Zilla i

1.1.la Maine Yankee FRA - Phase I Initiating Events Analysed 2.1.1b No ms1 Plant Trips j

i 2.1.1 Maine Yankee PRA - Phase ! Plant Systems Considered l

I 1.1.1 Rey Contributors to Core Damage Frequer.ty l

2.4.1 Plant Damage States 2.4.1 Plant Damage State Frequencies - Grouped Summary l

3.1 Phase ! Scope l

4.1.1 Types of Dependent Failure Encountered in Probabili6 tic 1

Risk Assessment j

4.1.2 Coverage of Dependent Failure Types 4.3.la LOCA types i

4. 3. lb -

LOCA Screening /crouping considerations l

4.3.1 Interfacing Systems LOCA Considerations I

4.3.3 General Transient Considerations j

4.3.4 Support System Degradation Considerations 4.3.5 Loss of Secondary Coolant Considerations 4.3.6 Initiating Event Grouping Considerations j

4.8.1 Abbreviations for the SETS Naming Convention i

4.9.1 A Sampling of Available Generic Data Sources f

4.9.1 Maine Yankee Plant - Specific Information Sources Containing PRA-Related Information 4.12.1 Useertainty Characterisation Approach j

4.12.1 Analysis Tasks Considered for Uncertainty Characterisation 4.12.3 Evaluation Processes 5.1 Reactivity Control

-vii-6606R h

l

]

LIBT.QF.TA211S l

(Continued) t i

i Mumber Title

)

5.2 Reactor Coolant System Inventory control l

5.3 Core cooling r

5.4 containment Integrity 6.1 Maine Yankee PRA Initiating Event Categories j

6.2 Susunary of Maine Yankee Initiating Fuent Frequencies 6.3 Maine Yankee Plant Trip Occurrences - January 1, 1974 Through January 1, 1987 6.4 Trip Characteristics l

6.5 Sunenary of Maine Yankee Occurrences 6.6 Sunenary of Grouped Maine Yankee Plant Occurrences 6.7 LOCA Mitigation Requirements f

6.8 Survey of PRA LOCA Trequencies

{

6.9 Nonnal Plant Trips j

6.10 Sununary of Important Actions / Impacts Following a Loss of DC Bus 1 That Results in a Plant Trip 6.11 Eummary of Important Actions / Impacts Following a Loss of DC Bus 3 That Results in a Plant Trip I

t 6.12 Recovery From Loss of DC1 and DC3 j

l 6.13 Sussnary of leportant Actions /lopacts Following a less

{

l of DC Bus 2 l

6.14 Sumanary of Important Actions /lopacts Following a less of DC Eus 4 l

7.1.1 Auxiliary (Support Systems) Event free Top Events l

7.1.2 Support Systems Event Tree Split Fractions

7. 2.1.

Phase I Frontline Mitigative Systes Listed by Safety Function 7.2.2 LOCA runctional Sue ess Criteria

-viii-6606R

i i

paf QF TJALES (Continued) i hahar Title

7. 2..%

LOCA Tree Top Event Definitions 1

7.2.4 14CA Event free Split Fractions 7.2.5 Traasient Tree Top Event Definitions 7.2.6 Trancient Event free Split Fractions 7.2.7 Secondary Sreaks and Excess Cooldown Event Tree Top Event Definitions i

7.2.8 fplit Fraction Secondary Breaks and Excess Cooldown Event Tree

]

7.2.9 SGTR Tree Top Event Definitions 7.2.10 SGTR Event Tree Split Fractions l

7.2.11 AWS Tree Top Event Definitions 7.2.12 AWS Event Tree Split Fractions 8.1.1 CondensateSystemQbntificationSummary 8.1.2 Main Feedwater System Quantification Summary l

8.1.3 faerlency/ Auxiliary Feedwater System Quantification Summary 8.1.4 Secondary Steam Relief Quantification Sammary f

8.1.5 secondary Isolation Quantification Summary I

8.2.1 HiSh Pressure Safety Injection Quantification Summary 8.2.2 Low Pressure Safety Injection Quantification Summary t

8.2.3 Containment Spray System Quantification Summary 8.2.4 BCCS Signals

[

8.3.1 Maine Yankee Electric Power Summary Results 8.4.1 Component Cooling Water Quantification Stammary 0.4.2 Service Water System Quantification Summary 8.5.1 Instrument Air System Quantification Summary f

8.5.2 EFCV Air System Quantification Summary 8.6.1 RCS Pressure Relief Quantification Summary

.ix-6606R

t u1T OF TABLES j

(Continued)

I hushar Title 8.7.1 Containment Isolation Quantification Summary 9.1 Maine Yansee FRA Plant Items and Associated Values 9.2 Maine Yankee Component Failure Rate Distributions 9.3 Maine Yankee Component Maintenance Frequency Distributions 9.4 Maine Yankee ' Component Maintenance Duration Distributions l

9.5 Maine Yankee Component Maintenance Unavailability 9.6 Yotal Number of Noncold Shutdown Calendar Ecurs for Each

]

Year in the Data Period (1974 Yhrough 1986) 10.1.1 Latent Buman Actions t

10.2.1 Dynamic Actions - Sequence Dependent 10.2.2 Dynamic Actions - Not Sequence Dependent

- 10.2.3 Initiating Event Recovery Actions l

10.3.1 Sequence Recovery 11.1 Containment CSF (Plant Damage States) f 12.1 Listing of All Phase I Initiating Events by Core Damage Frequency 12.2 Ranking of Initiating Event Contribution to Core Damage Frequency i

l' 12.3 Dual Vital Bus Losses Event Sequences

(

i 12.4 14CA lead Core Damage Event Sequences i

12.0 Loss cf Off-$ite Power Imad Core Damage Event Sequences l

l 12.6 8011 Lead Core Damage Event Sequences

(

12.7 Normal Plant Trip lead Core Damage Event Sequences 12.8 Imse of 4 kV Sus 6 Imad Core Damage Event Sequences 12.9 Loss of SCC Lead Core Damage Event Sequences

-x-6606R

.-~.,

l 1

List or TAmt.r.s (Continued)

Mudst 2111*

12.10 less of DC tus 1 or 3 14ad Core Damage Event Sequences j

\\

12.11 Damage States by Contairunent Mitigative Features 13.12 lead Plant Damage States 12.13 Lead Event Sequences for Dam se state SA 12.14 Lead Event Sequences for Damage State AA

]

12.15 toad Ivent sequences for Damage state 3D 12.16 Lead Event sequences for Damar.e state 5D 12.17 lead Event Sequences for Damage State 2A 1

12.18 Lead Event Sequences for Damage State 3L 12.19 land Event Sequences for Damage State 5B 4

i s

e

-xi-6606R c-

,e-

4 LIST OF FIGtf1ER j

Mumbar ZiLla 4.1-1 Block Diagram Structure of Risk Model j

4.1 1

1. i..

del.e,ments 4.3-1 Initiating Event Development

]

4.3-2 Initiating Event Categories

)

4.9-1 Maine Yankee Phase ! Data Base Development Process Flow Chart 1

4.9-2 Flow Diagram for Data Base Development - Plant Specific Data j

4.10-1 Example TEERP Tree j

)

4.10-2 Latent Human Error Application i

4.10-3 Flow Chart - Dynamic Human Actions

{

i

)

4.10-4 Operator Response Form 4.12-1 Incorporation of Parametric Uncertainties in PRA Evaluations j

5-1 Safety Functional Relationships f

7.1-1 Support Systems (Auxiliary) Event Tree

(

7.2-1 Frontline System Event Sequence Structure 7.2-2 Decision Points - LOCA Event Tree

?

7.2-3 Loss-of-Coolant Accidents Event Trees 7.2-4 Transient Event Tree

{

7.2-5 secondary Isolation Arrangement

7. 2-6 Loss of Secondary Coolant Event Tree 7.1-7 Stoes Generator Tube Rupture Event Tree

?

7.2-6 AWS Event Tree 8.1-1 Candensate system simplified Diagrsa 8.1-2 Main Feedwater System Simplified Diagram

-xii-6606R

(

-. -. ~., - -,

.~..-, -.. _,. -., - - - -., -..

7 7:x;,

c

~

• ~~

^ ^ ' ^ "

~

^~

~

~

^ ~ "

1

c. ) h,&,

.;,',., -.

.;eR!

' ' ^

,a 4_

..' ;'4 t if9: * *; '

g y Oy p1 m 7f; G k<M Q.

(Continued) j,j

,l s t

s 'Q,

+ ' t ' ; " \\' <.,

-,4

,,. x

.J " ', <

Manhaz 71L1a w

o.

,.,# 811-3; emergency and' Auxiliary Feedwater Systems Simplified v..,,t

' Diagram

~

.o V'.

1-8.14 Secondary Steam Relief Simplif10d Diagram

?

N 8.2-1 BCC8 Simplified Diagram a

D, 8.f f.

. Electric Power Distribution j

o,

8.3-2 Electric Power Distribution

. 8. 4-l' Cwponent Cooling Water

?

i 8.4-2 Fervice Water 9-1 Maine' Yankee (FRA) Pnase ! Data Base Development Process Flow Chart

.r.

E

. n t6

?

1 s

a-t W

e J.'

1'<

ii g

tf

[

1' t,

li s

5 i

e f'

li 11..

I l6

(

1:

t:

'i

.4 P'

t f-l;,c.

t l.'

,W l

i,' [

l.

l'

-xiii-l!

6606R i

n.uf.

'h 1

i

'm,,

~

4,

~,. - >

..f*

e f

1,' ', A.

_