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Forwards G Sih Comments Re Pressurized Thermal Shock Made at 811027 Meeting W/House Committee on Interior & Insular Affairs
	ML20040D116
Person / Time
Site:	Waterford
Issue date:	11/03/1981
From:	Roe J; NRC COMMISSION (OCM)
To:	Murley T; Office of Nuclear Reactor Regulation
Shared Package
ML20037D257	List: ML20037D256; ML20040D116; ML20042C227; ML20042C229;
References
	NUDOCS 8201300249
	Download: ML20040D116 (14)
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UNITED sT AT Es 2+

NUCLEAR REGULATORY CCf.'.f.*lSSION Enclosure J '

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/

November 3, 1981

...a carsct or int c M A18t M A N l

l l

NOTE TO:

Tom Murley Ja T oh ~

FFOM:

SUBJECT:

MEETING ON FRESSURIZED THERMAL SHOCX

~Cn October 27, 1981 several Commissioners' Technical Assistants met with staff members of the House Committee on Interior and Insular Affairs and Dr. Geonge Sih of Lehigh University.

Dr. 'Sih provided some draf t written remarks on t h e'rma l shock during the meeting.

I have attached these for your information.

~

D r.

Sih stated that he is willing to meet with.he NRC Staff to discuss his views.

cc:

Ed Abbott S.

D.

Liaw Tad Marsh Gary Zech John Austin Ed. Case EOy Woods D.

Ross M.

Vagins C. Serpan N.

Randall l

l t

i

.{

&l3$$ 2y7 CF

__.--------------_---.---J

uy c.

G. C. Sih Institute of Fracture and Solid I'.echanics Lehigh University Bethlehem, Pennsylvania 13015 I.

BACY. GROUND JNF,CRMATION This document is concerned with the evereccling ef-reactor vesseis and neutron flux embrittlement.

the progress'ive embrittlement of the material due ::

e c:mments are based on discussions with individuais fr = NRC and informati:n.

n

cntained in the referenced technical documents.

... Sina v.

nRT

.-- n 12, in Based on a numbe; qi_recent technical dccuments [1-3] cesiing with the in-fluence of thermai shock on nuclear reactor vessels, it bec:mes immediateiy evi-dcnt that there are great amounts of uncertainty associated with forecasting the Aside from the uncertainties associated with data collec-in e3.-ity of the RV.

tien en vessei fluence, irradiaticn temperature, e::., there is the uneasy feel-In a recent briefing to ing that the current methodology might nct be ade:uate.

the NRC -C:mmission [i] en Sep'tember 15, 1981, Dr. T. E. Murl ey - f rom NEC c.ves-s

.e p Oper criterien. Mersever, WF.at model sh uic be ti:ned whether RTND7 used to calculate RT

?

Indeed, if RT -

is to be _ ad:pted as an indicat:r for Nsi Nui the safe operation of RV, it then certainly shcuid be ider.tified with the mcde

c. taliure.

The utilities [2] view the threat of thermal shock a[s a icng-term problem tha: d es not recuire immediate action.

Eabcock & '.dic x (B&'.4), ':es tinghouse -

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..... c s,.,..

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be correlated with small ductile-behaving upper-shelf Charpy energies.

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ting under a different set of conditiens.

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Many of the structure failure prebiams in nuclear ;cwer piants are In.f act,- they were never understood and solved, and-are inherited frcm -

not new.

the ccnventional pcwer plant's.

Secause of the pctential health.harard associatec with nuclear pcwer plant failure, there is the urgent need to review the basic Ma n.y

-hilese.sh.v that uvides the overall design of the nuclear vessel structure.

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There is, in fac, no truth in any Of the pr:pesed #ailure' criteria ad pted by the ASME 5FV. Codes.-

Canais ancy in the basic c:ncert and method =1c;y, h: wever, be preserved so that knowledge and experimental data can be accumulate, d

have.t:

In this respeci, the present design philes-es t,ablish predictive capability.

Cur basic undsrstanding Only 50 years ago, TV and jumbo jets were unheard of.

T of science has fallen behind the rapid space of technoicey advancement.

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le.

cause neutron flux will interact intimately with material, icadinc. and :ccritr...

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t.

w..

a o

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b's. m.

4. 4 a

4....

J

3. =. *.*. { E. w. s

w.s a..s.

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.= g* E. 3.

g..

.g M a. g. m.;o

.g.4..g.. m, m.

0-.

3

l i

(

The material characteri:ation and stro:tural application aspects of nu-clear vess'ai design sh uld be e.phasi:ed not only in sciving specific te-hnical 1

pr:blems but also 'in establishing a s:und ;hiics:phy ef" design.

The nuclear ves-sei techn: logy is of sufficient nati:nai interest and deserves such a re :gni-

~

tien.

In this respect. a more critical review cf the ?GC len: ranc.e resear:h The ch: ice of failure er plan [5] ccccerning fracture mecha..ics is in crder.

fracture :riteria should not be taken lightly but must be carefully c:nsidered*,

i particularly for characterizing.ateriai damage behavice due to irradiatien.

4

]

Fcr instance, the strain. energy abscr;; ion to 'ailure [5] has been'shc n to be T

l' an effective criterien :: characteri e the embrittie. ent of metals in hostile envir:n.:en s such as radiation. The Hungarians [7] have also used this c:ncept i

for assessing the degfacnion of reactor vessel.cateriais caused by neut-On flux 1:;ingement.

l a r * *. *". *.*A' * "*.t*

t,r.

r'.

T: t:

a 1

Tj]

C:.. mission Sriefing en "?ressuri:ed Thermal-Shock of. Pressure Vesseis" by

~

T. E. Murley, September 15,'1951.

4

r. 21 Su. a r.y cf f'.eetings with P'r.'E C-ners' 'roup and Vend:r Representatives en 4

t uy e.... (././:.1 );,-e s t;ngnouse (//,../:.. ) an,.

n / <. :

4 "Pressuri:e,. inermat snoc(,,

+

i v.:3

e. -

1 1-q The mere 'a;reement bedeen theory and axperiment en a particular specimen er structural c: ;cnent is_ not sufficient.

Precictive :arability should be

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C:=ents and Ouestions on:

Comissi:n briefing (33 slides) by T. E. Murley on "Pressuri:ed Ther=al Sheck of Pressure Vessels", September 15, 1531 Slide No. 4 (Stress distribution)

I The pressure stress is not uniform.hrough the wall thickness.

o Slide No.10_ (Industry views - 3&W Plants)

What are the realistic assumstiens?

. (Oc: nee 1, Rancho Seco and t

What are the plant specific analyses planned?

~

l other, plants)?

(Industr.v views - Vestinc.h:use Piants) 5' Slide No. 11 a

All. Westinghouse p.lants can safely.. sustain. sever.e. thermal shock treatment, including repressuri:ation, t: beyond January 1953 -

t This statement must be based-en.some kind cf-f atigue analysis fcr predi: ting the life cf the vessel.

Are there any calculatiens available to. support this claim?

4 Slide No.12 (Industry views - CE Plants)

- Most limiting plants can safeiy sustain the most severe overcooling tran-sien

.,.cr an acditi:nal..- y

:rr. -

4 Are there calculations to support this statement?

Slide No.10. (Summary of Information Recuested from S Licensees)

.e,ine n:W,Di values fer we, ids!

v Mest welds when icaded are in the piastic range, i.e., the base material,

.. 'i.. ~>. a.. '. a l =.. d b.. a..=..= #. #. a... = d..... *..= - =. 2 1 1.v i a. i d. a. d.

W'. 2. d.a w e..= > r. '.v.. 4 'i s

.d ductility temperature in weids that are yielded?

f-ao y,..

3. : ( ;.. ~.., n.

4...:. w~..ca. d.

n T

-Oaio.

.j 0:es cladding -in f act enhance the creation cf smali cracks in-the base-A.

material because of the inte. face?

3.

On w.nat basis cces one :i nm t.ne parameters and mater 124 pr:perties are conserva-ive?

Slide Nc.. li (LOVT:5AjI Experience)

- M nitoring program shewed weld embrittlement. faster than expected -

5ased en what kind cf. test and what criterion?.

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o Comments cn:

Meeting with PWR Owners' Group and Vendor ?.epresentatives.

on Pressuri:ed Thermal Shock:

S&W (July 25, ige 1);

Westinghouse (July 29,1931);

and CE (July 30,1981)

Pace 1 (Summary of the Problem)

~

- Severe overcooling transients can produce high thermal stresses in vessel -

- Crack initiation can occur if a flaw exists and. if the fracture tiuchness has been reduced by neutron irradiation -

- Repressuri:ation duriag the transient can cause a crack :: p ro pa ga t'e'

~

threugh the wali -

- Someg?WR vessels are approaching high RTND* leve s where reduced fracture 7

toughness begins to be of concern -

t

~

Face 5_ (NRC C:~nclusicns)'

~

for continued cperation.

A limit should be established en RTND:

can be~ identified with the precise failure =0de in the vessel, Unless RTNDT wha t does a limi t on.RT DT ***" I" "I 5 U 5' 'IYI

,N Fage 10 (5&W Dsners' Plant Specific Program)

Are':hese calculations en heat conduction, fraccure mechanics, etc., avail-l able?

Face 44_(Fracture Mechanics Methods)

Are the feilcwing considerati ns included:

- non-self similar crack growth -

mixed m:de -

- material ncnhemogeneity due to nenhomoger. ecus irradiation effects --

- accumulative damage -

- c:rr:sien effects -

Pa:e c3 (Thermai and Stress Analysis) since K., is assumed to change with irradiation, d:es the other material 2.

pr:perties change ac'cordingly_or not?

' Face 60 H w-docycu. apply K..: to' the design of a ' vessel 1in a c mplex. stress state-An

-with La surfaci flaw?

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4.

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4

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1 l

7 x

' y C.7.r.ents en :

~

'l

,8,etter from Carrell G. Eisenhet (Director of Divisi:n of Licensinc NRC) to Henry D. Hukill, Vice President of l'etropolitan dis:n C:mpany

\\;

con:erning review of.he ?WF. Owner's Group on

" Pressurized Thermai to R?V",, August 21, 1931 (1). Cenclusion:

A11' plan'ts could survive a severe evereccling event for at

,y /.

least another year of full power operatien.

Acditi:r.al action

,should be taken new to resolve the long-ter=, problems.

New many times of ever:ccling during the year and what actic'ns i'

Cc=nents :

are beine. taken to rescive the icng-te-m problems?

(2)

Define'ET values for.weids and plates.

How does One identify these NDT values with the failure modes wi-hin the vessel structure?

(2)

'Jhat is th'e theoretical basis for estimating the rate increase of RibDT -

I (4)

The five cuestions raised in the letter. by Eisenhut sh:uld be.ans-ered with-in 50 days.

Are the answers available at this time?

f M

e 1

h f

e t

e

?

e 4

v

=

1 1

5 ENCLCSURE 5A

/'p *tCoq'o UNITED STATES' n

[

w,i NUCLEAR REGULATORY COMMISSICN WASHtNCTO N, C. c. 20555 s

. /

E

1M*,c -[<,

r

%,,,.. /

a., 0 4 ic:-2 r.

Darrell G. Eisenhut, Director, Division of Licensing, NRR

..EMORANCUM FOR:

Roger J. Mat; son, Direct:r, Divisi:n of Systems Integra icn, NRR FRCM:

TRANSMITTAL OF ': RAFT PALO VERCE AND CESSAR

SUBJECT:

nciesed are c::ies of our draft SER Supolements for Pale Verde and C..

yr,.

e December 1:,1:ci e

fhe'vaddress-heconcernsraisedbytheACRSle:ers,cinyciving the

,,h e lack of a direct means to rapidly depressuri:e the primary s i

bl eed me-hed of cooling the way it is provided in ether enRs.

Cur SER Sa:plements were prepared before the Ginna steam geneatae tub incident c' January 25, 1982 and represent Our position at

...a.

,1me.

The Ginna inciden has resulted in renewe'd' consideration being given t0 essibility of'simultanecus steam genera:cr tube ruptures in both sun,,c leac -

We are also recensidering accident scenaries that ::ut-5enerators.

inese h "ul7.necus loss of coolant in the primary and sec:ndary systems reexamine the ;cssibility of feed and bleed as

=

f

4. -

a+.4.ans rec.uire us t:

,s.e are a1so h..aras = 'pg ""*

~'1'". matned. cf :roviding core c:ol,.ng.

- ns an $1Uthe30RV to gain' control of primary system pressure to avold; c.]ai enges

- e safev valves en a faulted steam generater, :nere y reducins,.se use "freheney 5f releases cf radicactivity felicwing s:sa= generat:r tu e ru::ures.

a

~

f0 P lI O '"

.- re:arati n of Our Orlft SEE 32~0l CeUISma:1 n

,-y e,,,,,,",, y,ge)..'

... a c,. 4.

4. a.., s i nc =. *.*.a

- t--

anc C 55AR we have been provicec witn new inw.The new information 1s. in sys,.',,."

a-u13-ary Rase =rch.5 'ases its analysis :n the Acciden Secuence Precurs r er: gram.

.'.{i -ii4 v h~ a

' '. C...I..I.~....s s.,d in :hi s :r:c. ram-are sc=ewnat : nt-v ersial, 3r.d we Us E'

3e.c y3ngu.-

- ~er'i'v raviewing both

.e tecnniques and ::nclusiens.d January 2,:, l;c,1nw, i

i

'c

' v ? rank R ws:me and Jce Mur;ny of RES is date

- '.$'. d 2 d inc biset issue f r CE reac Or designs witncut r0./s.

-. C 2-

"'-"---".e an.-

..4":

,,e," eve,,"ese -.#.arancas u,

-.a k es we ::n.. a s i c,.,.s,._ nc... -,

53 3s.

variance with w rad, a

s stem wnich are 3:

uncer review.

C:ntact:

R. L: bel C 5.'. 5 3 i

6 AN6M Pr>s

~

i i

e-.a=

2-Carrell G. Eiser.hu In view of the c:ncerns discussed abcve we have evaluated the potential

nsequences o,f operation of San Cccfre Units 2 and 3 at icw pcwer for

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transfer ::

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= 4../. - 4.

a s..,..

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c

ATTACHMENT 1 SUPPLEMENTAL SAFETY E'/ALUAT!CN FOR CESSAR (SYSTEM 80) FDA AUXILIARY SYSTEMS 3 RANCH AC'RS CONCERN REGARDING RELIABILITY CF SHUTCO*4N HEAT REMOVA In the CESSAR Letter, the ACRS stated:

"In recent years, the availability of reliable shutdcwn heat removal capability for a wide range of t.ansients has been rec gni:ed to be of great importance to safety. The System 80 design does not include capability for rapid, direct depressuri:ation of the primary system or for any method of heat removal i= mediately. after shutdcwn which does In the present design, the not recuire use of the steam generators.

~

steam generators must be operated for heat removal after shutdcwn when This places the primary system is at high pressure and tem:erature.

extra imcor ance On tne reliao.lity of the auxiliary feedwater system used in connection with System 30 steam generat:rs and extra 2cuirements The ACRS believes that special on the integrity Of the steam generators.

attention snould be given to these matters in connection yith any plant g

The 00mmittee also believes that it encioying tne System 30 design.

may be useful :: give c0nsideration to tne potential for adding valves of a si:e to f acilitate rapid depressuri:stion of the System 50 primary allcw more director methocs of cecay heat removal.

c'ociant system ::

eview :nis matter furtner witn the c:cceratien The Committee wisnes t:

of C =bustica Engineering and the NRC Staff.'

3

'8 9

ee

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6

,ee h4m gg,

4 the c:ncern, the staff csition is presented In ceder. : fully res;ond t (1) auxiliary feedwatar system reliability, in three ; arts as foll ws:

(2) steam generat:r integrity and (3) the need for additicnal primary

~

facilitata direct rapid system dearessuri ation for system valves ::

.a.r.

. e a.......v. l.

. a. i '..s - * '. *..'y

..,,.s..... e n s'.,S "..~c =. e. ~.. '. a. r ' =. x. a ', - ~. c r..'..- =. ~....s.

.s SC Of :ne aux.iary feetwater system usac in ::njunc:'en af tn Systas steam genera :rs", we will re utre ina: Ocebustien Engineering incluce an auxiliary faecnatar system unavailability ac:aptance cri:eri:n as

,n 4..

4...

4.. ar.e.e.e. -n.3 w.e 33

4. 3. 4..s.

w..y.. f o. r..c.,.. ) <... a-4 The critarien will be One Oneir auxiliary feedwatar systam designs.

) Sectica identifiec in t.ie Stancard Review ?lan (NUREG-0500,

same as sna:

ng senerai Oesign Critaria 24, Residual Hea: Rer. val, 10.4.9 #:r.ee d and aa, 00: ling Watar as f licws:

m.

1..... s., i. :. 4.e $...u'a ' "..aV.

.s n... s v g...s '. i. y..a.

..w. a.- s... : o.1. 0

.a d

d*

2

i. o.

m. e d.....s..*.

. a t..s e.

. = =. s n.s.a..s i.v. J. s..s.. e...a.

6.a.e. s n.e.c d.3.3

a. o.s a.. s s.

4..

i.,.

v... e t..v- :...:.

...,. s.i. e.. C 23,.

7 2...s....

,..e...Cs

.... 2 yn s..:.v : *6

w. e

4. 5

...e e'"We.

P....e e.

.s.14. s.a. i s.

..e....- d a-

2. e..y

3. ". ~-.d.

n s a d.

r a3 3....... i t.. 4..,.s : s 2

a...

..s-..,

-.n.o.

. o. 4... n s n e... 3 ) ~ ~..gi.4. c n g

.3, m e..- e. g.. a. e.s.

y...

t..;

s..e..cl..-

.. 4.. $. J. J..y a

t.3.-..

. s.y 3 4. *i s

4. ] 4..y

..a. } {y.t. *

2 9

7

i We conclude that this interface adequately addresses auxiliary feedwater system reliability for CESSAR reference plants.

In regard 'to the ACRS concern for " extra recuirements en the integrity of the steam generators", the fellcwing is the staff position.

The System 50 steam generators incorporata multiple design features to minimize the instance of problems wnich have been identified to date These features inicude ' improve-in operating plants steam generators.

We note ments in material of construction and fabrication techniques.

that tnere is no operating experience associated with the Sytem 30 steam generators. Therefore, we knew of no reason to impose additional recuire-ments at this time for assuring their integrity.- If operating experience indicates that additional requirements are warranted, we vill incorporate them as necessary.

It shculd also be noted that the CESSAR SER (NUREG-0852) incluces dis and staff conclusion en steam generator integrity and certain aspects of steam ' generator performance as fellcws:

Materials and fabrication and their acceptability agains*

(a) acplicable ASME Codes and General Design Criteria are addressed in SE?. Section 5.a.2.1.

4

+

.mM use me egia

=e6 e. ge mswe

-m,

H Design features for 'preventien of damaging Water hammer

. {b) is addressed in SIR See:fon 10.4 Secondary water chemistry is addressed in SER Sec-icn 10.3.1.

(c)

Basec en :na 1:cve, e ::nclude tha-the in agri y cf ne Sys am 30 s:aam generat:rs is acecuate to assure -tef e availa:fif ty f:r decay nes: remcyal and that furtner recuirements in :nis area are no:

neces sary.

4,,

e..

e..a..n..rn. r

...s.s...,.4.

n....,,............

4 s.

~....

.... gar......

s'e,-

s f

.a d..' '.,. g v al v a.s. #.

a s'.-a. 'a

..$ 4.1 1.d..a..a

-s. a. i.d d,.a c r,as 2. ri..=.d..a, a.'...a

-~

s decay nes 01 ant systam Oc 111:w more cirect metneds :#

Sys em 30 c:

removal,* :ne #c11: wing is the staff ;csition in this mattar.

....(.. r.L,.,3 ans. a.

v

....a.-

aj....a.s..

. e... c

d o r..s.y se

... ~

~

...r.s.o.r....

aa..,........

4..

.......a.

.o s w-..... 2

.V...

l*.

4.

. a....

.e s..sm 4... 4 3 4.

a.

4......

...ev a*t Na,

...n.

a...

...s

.y s.;.,..

.s. 1:..o.

4..

.g.

-.4s......

.e.

a.,, y

..mw 2.... s.... $

2....

.i..a.,

4 ym.i.j c

...]

3.

4.

4. n.* **e *,*.#.a.v.

sy s *.*7.. V. *4 **e 3..

t. 8. # V 2. '. V a. s.

u.p *.....s.'..*

*. '. ' '...4. a*.- '. a *. ": a. 1. 1

2. '. ".. *. *. s.* #. *. *.*/

i

.y.......=.4....... e

.a s '.... a..

3.- e...w 4.$ o.r..

3 3.

....34

3.. y

.y y *.% r.

..p,. }.s. 4... e.

.e

=

H $

M b

m.

m k

Q

l 1

i In order for feed and bleed to be avviable decay heat removal mech tity of the HPI system must be capable of injecting a sufficient quan coolant at the prevailing system pressures.

V For plants without a manual depressuri:ation capability (i.e., PCR s/ stem with encugn relief ca:acility to sufficiently ce;rassuri:e the loss of' primary system), the crevailing system pressure folicwing a 0 psi).

all feedwater will be the safety valve set pressure (usually 250 Thus, in order to have a viable feed and bleed capability in plants i

fficient without PCRVs, the HPI pumos must be ca;able of inject ng su This implies quantities of coolant at the safety valve set pressure.

fety the need for an HPi pumo chut0ff heat censiderably aseve the sa valve set pressure.

f Fcr plants with HPI pum:s that de not have shut ff heads abcVe the s i

y system valve set pressge, a means :: manually depressuri:e the pr mar r

in an to a ;ressure sufficiently :elew the HPI pump shutoff pressure a ectable accunt of time aculd be necessary.

i tion PCRVs wcuid typically be relied d;cn to previde this man'.31 depress for' via:le 'feec and bleed" ca: ability.

9?

CE SYSTEM 30 CESIGN The present C ::ustion Engineering (System 30) standard ;lan; des The HPI system does not include pcwer-ccerated relief valves (PCRVs).

Thus, in tne emolcys the cum:s ith a shutoff ;ressure Of 17s0 psig.

5 6 h N

ma O

'M@&4 e

d event of a icss Of.111 fe'edwatar, the System 30 design does act have

-he capability to de ressuri:e -he primary system to bei:w -he HPI Thus, in inis design, reliance cannc: he placed en shut ff pressure.

"feec and bleed" for decay heat remcval.

t.

e. iA: :. :. r..e.i. i]

.= n d ". I =. a.d.

i.

' ' h D. e

.e s..= '. '.. a.. :,. i.- a s...e c..=.n.i al

'.e.n e #. 8..s

..a a.d.

.-a. -. a.s =...l y-

.a

.d.a si 7, r a.,".i - a.m. e....-... '...= r..= a r. i..,.

d

.. a

4. '. '..v,

.*.a. r a.

y

.. s culs re uire CE system 30 plants : install an alternata decay hea:

The staff remeval system indecencent Of ne staam generat:r s.ystem.

... s. a.. -, r. d. - a.d.

.a.

.. a. a. d. '.. r -a.l '..* 'i e e.

y-

a. =. *. - a.. c y '..

.'..e 2.a#. #.

~

acce: ance :ritarien for auxiliary feecwa te system ( AF'S) reif acility (as iten-ified in SEP Sect'sn 10.1.9) is based n an ac:e::ance Of :ne

..e a.. v a.l ".a.

. '..". a.,r.'a-fli.v. - '...ra. -... e l *. '...,. '. a a..w a *..*

- *.. s t a.... s Sna was derivec in WASH-iaCC. The s aff rec:gni:es One ifmitz-i:ns

~

in WASH.'aCC as delinea ad in ;revi:us sta aments. Mcwever, in using

... s... '.,

a.

-.e= v a.

$ k a a.

..e.>., l '..= o 'i e.-... a. n. e.n.. a.-.

...7 '.a s.=.

i.

s.e. e. l. s e.

.s -o.nse1.I Qf g a a n.o..t..a

..s

s m a g a.

. a *

.".e.. e. e. n. s '...-... g T. g. n ga..d 4

3 w

d.

s.

.s...s e. 3..s ". *... s.e.

s..r. d.-.. e

a. 4. J... ],v

..a A e..s r. 4... s.

,a..sk.a.saas q

..g

..e a '. #. #. 8. #. ~ #.

1. *.

w 4..

1

~~

2...

4. 5...,

,s.,.r.

3

.e -

e. 3., 4,...i..., -.... 3. e.

.3 3,.

j,..,.

. w.. g g.

4... '.,..,. 3 3.,....... a.. y

.a....

4. 4.y

4..

2 s., t. e., s.... t. e..,, a.. s....,. s.. e. S t a

.t..,C.... a.....t...t. a.,. s..,u.t.. s t..

..w

..,.4...g......: :s t..

4. 2. t.,. -. 2. a.

.4...,4.. w3,w..

5

.e 1

3... w.

au.

.......,y a.

'I s. *. *. a.

'.~..~.a 2 2 =. a =. t.,s s =..am.n.

- s e. '. : ~.. -."a 3 4..- s e,$ s.c ?.

.s

n'4. e a.s =.4. 2. t. a.J.

a e

E.

9 e

---g.

y 9._

Addi tional

. by all future System 30 plants (refer to Part 1 above).

mitigating features available to satisfy the core telt risk probability This is discussed further would be evaluated on a plant specific basis.

in.,he Palo Verde SER Supplement addressing similar ACRS concerns.

t Notwitnstanding tne present reliability requirements for AW systems and overall decay heat removal capability, the staff has initiated work on the unresolved safety issue of decay heat receval reliability (USI A key element of this prcgram will be an evaluation of risk A 45).

reduction that would be af for.ded by a viable " feed and bleed" capability.

If it is concluded inat a cost beneficial reduction in risk c0uld be achieved by incorporating a " feed and bleed" capability in cperating plants that presently do not have such a capability, then a backf't order aculd be considered.

Mcwever, until this study is ccmpleted, the staff concludes there is no need to require a " feed and bleed" capability be installed in System 30 plants since acequate heat remeval system reliability will be assured by the 4RS reliacility criterion as an interface requirement in CESSAR.

It is :ne staff positien that the present AW reliability criterion Meeting this must ce met by a:olic' ants of the CE System 80 design.

csition Orcyices a sufficiently 10w probacility of core melt for this design, and further assures a reliable decay heat removal capabDity.

In summary, we conclude inat the CESSAR System 30 design for decay heat apolicable 3eneral Design Criteria and guidance removal conforms ::

and is sufficiently relia:1e :: assure safe shutdcwn.

7

1

(

e 4

e 1

l i

e L

t

. 2._ :. _

ATTACHMENT 2,

SUPPLEMENTAL SAFETY EVALUATICN FOR PALO VERCE NUCLEAR GENERATING STATION, UNITS 1, 2 AND 3 AUXILIARY SYSTEMS SRANCH ACRS CONCERN REGARDING RELIABILITY CF SHUTDC'WN HEAT RE. OVAL SYSTEM M

In the Pale Verde letter, the ACRS stated:

"In the Palo Verde design the primary system coes not include

. -capability for rapid, direct depressuri:ation when the plant has been shut down. This places extra importance en the reliability of the auxiliary feedwater system and makes it necessary that the NRC Staff and the Applicant assure the availability and dependability of this system for a wide variety of transients. It also places extra recuirements on the contintied integrity of the two steam genera.crs as the only me'thod of heat receval immediately af ter shutdcwn.

The ACRS recc= mends that the NEC Staff and the Ari2cna Public Service Ccmpany give additional attention to the matter of shutdcwn heat removal for Palc Verde and develec a detailed evaluation and justification for the pcsition judged to be The Cc=mittee wishes to be kact informed."

accectacle.

The follcwing is the staff ;csition en the abcVe concern.

In regard to the ACRS ccncern for extra importance placed en the relianility of t.e AF45 in view of tne lack of a 9

racid, direct de;ressurizatica cacacility for the primary

-e a

pw e,

N.-. _ _ -,,

. ~..

system, and the A"RS ree ::endati:n for a detsfled evaluation

'.e

.$..*. *, *..= b l e,

and i u s

d. '. i c a *.d. a. n '. a.r.' e a.o s'. *.'.o n

.d u d e.a.d.

a the folicwing is the staff ;csition :n this = attar.

eus:.4-ve-:.),nd r...m.r... 1.1 r

s n

. =.e

r. 4.. n 3.e. c f... o 1. sI o..a...

e.r.R a

a.

e.

. ~.

...e '. a..

..*.... e.=..1 3.

a e..av. e. s...4. :. 4...

w.

. e....e

. v.. r. 2. :..,a$., em e e...,

su:mittad an AFWS reliacility stu:y in ac: r:ance wita staf f guicance.

One AFWS me One system

he staff reviewed the study and deter:'ned ina

' t: 10

er demand) f: r a unavaiiacility ac:a::ance criterien (10 1:ss of all feetwatar as & result of a feedwater transient or loss We al s: detarmined :na :ne AFh*S desi.-n

f #iste ::aer initia-i g events.

t.

e,.z.,n io.1.a.

s. _.e., s.a,.s. a v.,w

.<.,r.1 s.

a

.e...

,,, 13

...,rmt.n;s.t..

i....

r N.J....:n0.

.lan s

2-

n aedition, as the AFWS unavailability acce::ance criteri:n is derived

=0 fr:m a r'sk

::re melt frecuency of 5 x 10

er rea: ::r year ("eact:r Safety 5:: y, WASH-iaCO) ::nsi:er$ icn was given :: addi-icnal pl an:

" ' ' s *..S 7.. '.'7 **. I #. * *. #. l i *.'/

wp....4.;... e.. s...q..

.. E s*.~9' '.

2.,

s y s..., j.

4*

4..

.1

,.5

a l. '. *. a..".=. a...v 3.-....s.-=.

.. '. '. a. d. e n "i.

.- *0-

a. a. r.d em...= a...- ).... *. -. r =. u.

t w.

~%..g g e 3 j a.d a *.#.."."3#. 8. 8 *". *.* s

..".CI ".d.*

a s'..' h 'i *.

-c t

63 f.: x

o.

. a. *. _ e. a.g =..= *..v a. s =. ).

. s i M a. a.. a e..s--

w - d. l *.*. v.

d. ~ a. (.$ "... *. *^ d. *~ ~* *.*. l.v. " ".

'.#. " ". *..* s i m

=d,-.

sa.. r. *..:

s..

g.

#. #

A. *.1

.*.W e P.

a " ".. l.V g m.. d.. =. 16* *.9 s 4 a. *.

a. m p t. * * *.

e.e. a.*/ O. P. '.s. ". b. e.

- d..*

2.*. F.

-2 3

"

8 2.*

.d.*.S. *.** ". g ia 1* *1...*

V a. *. d. e.

. $ *. *.*....e E.*

19 'I a.

J.

=

= =r ed

i m a.

1 *

1 a. a s.a... m. e.*.

v a.a... p s. e.4..-s 3 **c.3

=. J. a #. #.

4

s a

di..

w. m.

. i.t.*. *. J e

". b.b* *4,

- 4 E a s. e.. e..a. s.,y a

.Y m

6

a

'~

---g h

'6

i i

t

+

s pcwer should be equivalent to the average assumed in past analyses, Further, the 20 minutes approximately 0.2 to 0.4 per reactor year.

of steam generator water inventory af ter a loss of main feedwater allcws time for plant operators to restore the AFWS should it fatl initially, Previous estimates incicate or restore offsite pcwer and main feedwater.

approximataly a 40% chance of restoring offsite pcwer within 20 minutes.

These features provide additional confidence that the risk of ccre melt

-6 is not exceeded for an extandec 1 css of feedwater probability of 5 x 10 condi tion.

Based on the above, we conclude that the Palo ierde AFWS meets tne staff reliability acceptance criterion and further that it is unlikely that

-6 will be exceeded as a the risk of ccre melt probability of 5 x 10 result of feedwater transients.

In regard tc the ACRS concern for " extra requirements-on continued integrity of the two steam generators as the only methcd cf heat removal immediately af ter shutdcwn," the folicwing is the staff position.

The integrity of the System 50 steam generat:rs has been reviewed by tne staff ar,d found to be acceptable. Refer to the CESSAR SER Supplement Further, the Palo Verde SER addressing ACES concerns on this subject.

(NURE3-0857} includes discussicn en the acca:tability of the folicwing relative to steam generator integrity:

5.-

M

,g m

6 ea' %r #*

.* e g gg g

e

1 The s eas generat:r insenice ins:ec:1:n ::r: gram is 3) addressed in SIR Secti:n 5.1.2.1; l

b) 'The see:ndary watar chemistry :: nit: ring and c:ntro

r: gram is addressed in SER See:icn 10.3.3; and

/feecwatar 3ree:erati:nal testing f:r staan ;eners=r

)

tica

aterna=er preventien is addressed in SER Sec

,o..

s

.o.

tne ?al: 'ler:e s el ;eners=rs Oreclu e Oa:

Basec :n ne ab ve, we :

-e 1e :eans #-r shu ::wn decay heat ree val *in:c:

e

-7.,vi ce a -ti f a:

inuec integrity.

neec #:r aceiti nal requirements for assuring their ::nt r

hea: receval cap-n summary, we c nclude P.at :ne Pal: Verde shute:wn q = "1 c.1 4..,31,..,.a...-........s

.., a,,,,if..'.*.

J=..

f 14,.y

4. 3 3~.e :. 4..4..r..1,v Oesign ;riteria and guidance witreut further requiremen 3.-

e 9

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h. g N

ATTAC:ll'E?iT 3 UNif t O al Atl.$

' ' ha' ' ',

me.tE AR HCCtjl.A10itY CO*.**.*1MION

s.

~

a WASHINGTON O. C. 20G55 2 *.*

5.x:,;g;,1 o

e,b,.t;a,/*f JAN 2 71982 4

~

NE:CP$10U!!FOR: Ecb Tedesco, Assistant Director for Licensing Division of Licensing, tiRR

(

1 l

Themis Speis, Assistant Director f0.r Reactor Safety Division of Systzns Integration, fiRR FR0ti:

Frank H. Rouseme, Oe::uty Direct:,r Division of Risk Analysis, RES Joseph A. Murphy Reactor Risk Branch Division of Risk Analysis, RES

SUBJECT:

FEED A 10 SLEED ISSUE FOR CE AFFLICAtlTS 1

We have perfenned a quick and dirty analysis of the risk implications of CE designs that lack a capability for core c:oling via HPI injection and deliberate ventihg of' the react:r coolant system, in the absence of feedwater replenish ent.

than

'a'e c:nclude that three classes of accidents may each he rcre frequen:

the Commission's safety goal of 10-4 core melts per react:r year or less, and that the total core melt frequency for such plants c:uld be of the Order tf 10~3 per year Or ecre. ~he three sequences are:

Transient and failure Of all feedwater (not associa 2d wi.h ioss cf l.

AC pcuer) (TML).

L:ss of offsite power, one diesel failure disabling the mtor driven E.

AF,1 train, and failure af the turbine-driven AFW train.

'lery small LCCA and failure of HPI (S{D).

3.

4 W h oS~q y

.___.-,o.=*-~~-

..-c

f

.ic.: .. wend the following upgrades to thesa designs-d Provide an assured "Tecd nd hiced" capability.

1.

Previde that aither dicscl genera:ce can energi:e a r.c:or driven AFW 2.

.,4n.

.- u... =

p,..,., 3.,., s.v. s,., s..,, s J n,, r : d. =. '...: *. r s. 'i '. 2 ' '. 's '. v a nd i -

the frecuency cf very suall LOCA's.

The ec:r.ciaic inc:ncives :: ::aka chese imprevements, derived f-::n r:duced

.4,"...... *.

..e '1. a,.= re... u.5 '4 v -

4 g

,<eg.e.

e.r.. c..1 '. c '.. e.

.= s = ". c '..= '..a d.

2 t

Base Case.

'/alue 2:a. M

,a. e 242..M

. ic

\\

Y

/

3ase "ase with 3c:h 3ase 'ase wich G's

.;1'.-.s..

.a s eu ro.a.

i, s.:q i.c. r., r', v........-

n g,

.s o..

... : ~. a. u..

e ~s., -.s.

,I., <t...

.v.c

.e.,....

.e::- n.e0 I*

/

I Assursc.eec.and312ed}

2 CG's - 2'AFW Trains;

.e :M o....

V

,issured Feed 1. d 31ee.d ll i

,.-. ra ns

...s-

...4

.a I

l :-!igh.:eliabili y HP' I

a e

4

.4 6# 6

, og g

g

,g gg#

,,d a

g

~

The base case plant is assu=ed to be incapable of feed and bleed ccoling, only one diesel generator is assu;ned capable of energizing the safety related..ntor The turbine dr'ien arf train is AC-independent, but the _

driven AEi tra,in.

non-safety grade motor-driven AF train requires offsite power.

Industry The analysis that average Phliability and 3 -LCCA frequency is assumed.

2 shows that S 0 may be too frequent applies to other p'.'Rs as well.

2 The attached paper describes the analysis.

/

A

- W J V&,.-.~ _

Frank H. Rowseme, Deputy Director Division of Risk Analysis Cffice of,iuclear Reg'.latory Research

}'Y ~

+f' Joseph A. Iturphy Reactor Risk Branch Division of Risk Analysis Cffice of Nuclear Regulatory Research

Attachment:

As Stated cc:

R. Bernero G. 3urdick '

R. Mattson S. Hanauer M. Ernst A. Thadani RR3 Staff RAS Staff 1

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J Feed and Ulcod Issue for CE Applicants

!!c understand that the current crop of CE license applicants are proposin i

that no pressurizer PORV's be installed, that the HPI shutoff head is to be well below the pressurizer safety valve setpoint (around 14C0 psi), tha high point vents provide no more than two 1" diameter rer.ote-manual and that the auxiliary feedwater systems will be c:cposed of cne AC-in turbine defven pump, one AC-power train, and a third non-safety grade m driven pump.

We have attempted a back-of-the-envelope FRA in order to evaluate th implications if these plants are incapable of " feed and bleed" cociing.

The results suggest that they cay fail to meet the Commission's safety of a core nelt frequency less than 10-#/ year and the present worth of a i

to enable assured feed anc bleed cooling is of the order of $10 :ill en or Ne c nsidered five more per plant, based upon reduced financial risk alone.

loss of main feedwater, ioss of offsite pcwer, groups of accident sequences:

d very small !.CCA, transient-inhced small LCCA (late start of auxiliary

)

water allows a lift of a pressurizer c:de safety valve which may stick open,

and station black:ut with restoration of AC pcwer just before the point-o return. We did not consider main stJam line breaks or ATdS, althcugh in

' seq 2ences'an assured feed and bleed capability c:uld also enhance sa

-well as in the sequences considered.

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this sequence in a plant incapable of feed and bleed ceci'ng, the frequency 4

h s the frequency of critical (sus ained) i Of c:re melt,1, = 1, P(L}, where i i s

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he 3 er year, with

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ck the frequency of #eedwater transients ::

There is reasen to

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Data frcm the precurser Auxiliary feedwater reliability is also uncertain.

program suggests that the FWR average experience has been a fai,1ure prcbability This aierage includes early-in-life experience as well as of 10-3/ demand.

mature plant experience and two train as well as three train experience.

System reliability analyses have suggested tha-the best of the three train systems can apprcach - at maturity 3 per demand. Mcwever, these analyses failed to censider sc=e ccamen ecde failure mechanisms so they, can be regarded It is. net unce::en early in plant life to find as having an cptimistic bias.

instances of repeated, consistent, auxiliary feedwater pump failures while The record suggests that the failure the system is being debugged in service.

probability of the AFWS is substantially higher during the first ccre than in A system with two diverse safety grade AFW trains and a third full

na turity.

capacity ncn-safety grade train will pecbably achieve failure pecbabilities of:

3 x 10-3d, first core r(L)=

3 1 x 10"1, at maturity These estimates result in Icss-cf-all-feedwater frequencies ef:

0.9 x 10 ul*'/yr, first core 1 x 10- d' /yr, at maturity I

C The uncertainty range is thus:

2.3 x 10-2 4 1 a & 3.5 x 10", firse core c

2.5 x 10~ t x

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Assume for c:nyentence that diesel generator A is configured to energi:e the safety grade AFW motor driven train. As we shall see, the ccre melt frSquency predictions are sensitive to whether or not diesel generat:r S The event tree for can energi:e the non-safety grace AFW train or not.

loss of offsita pcwer can be drawn:

CG's AFW

--) :kay no failures 10-4 3 melt at 4 x 10~ /yr 93

--4 okay B falls 10~3

> :elt at -1.2 x10~5/yr

.03 LOSP C4

--> okay

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,--y okay, -

both fail

):elt at 1.2 x 10-5/yr

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The higher failure rate applies if. ne of the diesel generat:rs (we have called it 3) : anne :.cwer a met:r dr.iven AFW train; the icwer failure. rate applies if both diesei generat:rs can pcwer a 20t:r driven AFW train.

4 Meta that the C:=missien safety goal of 10/yr for all ::re melt sequences may i

te vi:ia ad by icss Of offsite 70wer and a single diesel generator failure if energi:e a ::t:r-driven there is cne diesel generat:r that cannot be aligned ::

This high c:re melt frequency c:uid be reduced t: marginally AFW train.

.acce::able value in,either of two ways:

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iding a swing bus f:r the safet/

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i h this feed.en:.:Ised fix Can be infarred e :m the event tree f:r L r

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In th CE plants, botn feedwater and ECCS (HFI) are required for successful cere c cling. Main feedwater may remain operable or be restartable in sc e of these. The prehbility of HPI failure en demand was f und to be 3.5 x E

10'3i' in Surry (WASH-1400). Most PWR ? ras are finding a failure pr:bability for the whole multi-train HPI between 10- and 10-3/ demand. We.shall assu that the probability of HPI failure en demand is 5 x IO~ d/ demand for the CE plants. A rough cut at frequency estimatien suggests:

HPI AFM NFW

' suc ess 10- E

}suc;ess

. a x 10 5 S LOCA 2

1, el. a.

3 x 10* '10-71I /yr x 10 uI 1 :eit at 1.5 x 10 a+1*1/yr The value of an assured feed and bleed capability here is t: eliminate the need This wculd eliminate the smaller (10-6/yr) path to core melt for feedwater.

wi heut affecting the =cre prc inen: path via HPI failure. Note that small LOCA w th :tal HPI failure is predicted to resuit in a c:re melt frequency above the C::.issicn geal for ali :re melts.

ine prev sien of feed and bleed capabi-lity er cf an impr ved AFW systam will no: help this.

It is a pr:ble: generic

PWRs anc an: unicue :: the CE designs. It appears that the high frequency of very small LOCA revealed by his::rical experience and the marginal H?I system reliacilities revealed by many ?'4R ? ras are c::bining to yield unacceptable core l

melt frequencies thr ugh S C-type sequences. We suggest that F '.R tackie this l

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e i

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reduce the sericus effort should b;e made to a: tack en HPI reliability First, a ys:

reb ~em in tac wa Sectnd, a bread-scaled fer A?W systems ld be LCCA's.

frequency of Sg

hat ins-i:u:e

re: ie:s c:::arabie ::

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-e 4.4. 4..,..a* s..

ed small LCCA's, wi:n and withcu I6 s 6.

ist us ::nsider the transient in uc aut: start of auxfliary feedwater d

A feedwater transient with a prc=;;

Mcwever, a deisyed start

'.lext r relief valve.

a sustained AFW F0F.V.

is assumed nc: :: itft a pressur :ehly ene hundred t mes i

lve may Of AFW, wnich czy be roug l e (?CRV cr c de safety) and the va failure, may lif t a pressuri:er va v v...ilanges w,

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i t< m 1 css of all feedwater has already been censidered.

The core melt.outc:r.

LCCA is negligiole at 10'G

/yr.

It can

.The increment in tb likelihood of 52 jority of still be mitigated by HPI, if HPI works, as it will do in the vast ma i

cases.

(10-5/yr) f With a FCRV we will get transient-induced, LOCA ten times as e ten, t of these but the bicek valve can be expected ~to terminate'all but 1 percen

-7 /yr.

If for a frequency of transient-induced and unisolated LOCA cf 10 ible anything, the.FCRV helps rather than aggravates what is a neglig 7

frequency via transient-induced LOCA.

c:ntributor to the overall S Z

icus "c;en

We shculd aise censider the ccmmand fault LOCA's due to spur The frequency of cccurrence is a sensitive function of c:=: ands to a FCRY.

It could be made as small as we wish by the valve c:ntrol icgic design.

If we censider the Crystal River experience suitable reliability engineering.

7 3x10-3/yr f:r as One failure in 300 FWR-years, we get an industry average 0 Clearly, B&W did not do so well, but the c:=bined I

PCRV'cpmmand fault LCCA.

cy can easily ex;erienca cf :he three PWR venders suggests : hat this frequen 3x 0 'It fyr.

I c:nclude requency e be*:ade much'less thtn the Overali 5 2

i ible effect en the likeliheed that having a FCRV cr not having a PCRV has a negl g LCCA' may lead to c:re melt, previded LCCA or of the-likelthecd that S g

It 2

h ly censideraticn.

of 5 that system or c:mponent functi:nal reliability is t e oni with antici -

gces with:ut saying that this analysis is predicated upcn a des gn t lift pressuri:er relief valve:,

_ pat:ry trips se that reutine transients do no

nenappr: priate.

S l

and that the'eperat:rs are trained t: clcse : e FORY bicek va ve w

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10 I feel unc:mfortable with lac 0 psi There -ay also be a design adequacy issue.

I and the AI'n systems MFI pum:s in piants with0ut ;CRV's, even if the HF Carefai ther ai hydraulic analyses t:gether with are afgnly reliable.

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HF* sne AFi rains.

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a frequeney of roug gdekcutwitheeauxiMaryie an be expect' g _of.nc-return M N

t.me M wgp te high.

4 "ither Scenado' 15 the..m pre 550Te the cdf8 Y l wuI D*

5x10"*/YT-react 0f COC lve set 90i ")' and,he leve 33 generator 5

+ 5* e cacne5.

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dins pon 'he fh u

A eg u ng t0**fd ""e tC9 not be Sufficie#' depen SYS. m TeakaSe-wi u i.e necessary "ut mayexte" of reactor eciant. p.*. reac*c r

svstem conde W,cn and 'he 8

of reflux o enable R?1 ^

without core damaSe n eu f r Ac rec W yitative.vDuation of feed and bleed capabi M ex'and tbe,

~

fiirTY qdcW

nnu,. 5' "erhaps 50f**

. *d Lnd b,eed wcM Codd be 51Vid "~Y "'"'

e g ged hecd of AC t $ ' "

melt DY fraction Cf g: sequence 5 '* hat ngysis of

,.,o ggy Ac re5

ration r

an U *,ngysis t* is Ciear tna "e ;CS,*

g en.

4.F" revECeu'

.gtensiV. 'h' ermal hY r1U d

Mcwevef'

_hC5'

.n upper --

dwe. disc,h

-re5,.,wfggion V5 4_e*

a t

r e...rn,

dg -

any 7 gg of no cy a.. 4' utable g f g d an 4....es are tefora 9uence..quen e..

N..,e diac.sou.."a s,

E?l 6/Y,', or ie55 C; s signs 'd*' *h ICW ef 10

.he ceder eca,i~'ng :ne

,y cerng r -

su Cat he' "w. princtpa3 c:n be*

.a US P

D3y high.

shut 0,',' dead and nc ?OE

.,,gd.gater ~1Y ~. unac.a?ta w~

,. yta tC55 0 a$ questiOUAb 40 a 4 s k of ~ 0Ii,,,',", '

very 5_"" D 1 CCo* '"iticat 0 1

2-n for adequ1CY

.y berader issues-I ~h" ne O y...pged wi 5 C;eri

1'

c..

Q {s maY

,,a

12 3.

The reifability of the high pressure infection system ay be unacceptably Icw, but the mere fact of an AFW requirement't: mitigate very small

, L^CA's. given design adequacy. dces not significantiy degrade the r= liability wi:n which very small LCCA's may be mitigated.

4.

i s '... a.c e..' n. *..'.a.

a. f e..".e r a i s.s e *t

,.a.n a..'.. r '. e..t

.=.41 = '. =.. a.. d.. '...,

. s z

.:, s. zs t... r.ae.

a.c... a. r../.n.s.v..., 4. n 3../,.n. s a.

a i.

Two questions remain :: he ansaered:

(1)

-3: is it acr:h t: equip these

,4 s. s

.*3.s d..w e...s..ino w*..s.

1*1..f..

nd g.'I9) ah. a..

... s....a.s 4...

a 4

..................i<

..r..

.:. s.

4... s *t

e. 4. *Xs,.

.ms 3ss s(.n'.

J.

. h. e V.3 I o..

m J..a..

84

. a } 's.a Ws.

i.=.C s e w re.

a...s 's *. 3 C a. 4..J a. *.

s. a n. o..m. a. s aa ia w.

ep...

-.5.s

e. a. d. 2.a..

l *. *. d.

~.'. a ' #. i. d..,'/.-.u i ;. a.s./ a.

a..b.. - a.. s.

2..s.

1 3. *< a '.,y

.0

2

- A 6

33.'i

,.......C n..- d. :a i *i,..... r..w.a 1,...s

.a n 4. a..a u *i.

s

...si.

.s..,.w.y s..e a ':

.4 i........

,.w.

w..

. a

.s.e.s.a....... 3 4.

.. g. j e

?..s. 4. n,

a....,.j g, 7 s..s.,....*

.J. a. i..., s.

4.,.a s

r..

..6....

u$

.w.

.....4..2is4

4. a.,e.

9

b. d. *i a. a. a. (' *i.a.w.

.3

.. s *..n < s * * *. a. a s =. *. J.

. ". a.h.

a w= a.

s...i*.

3. v a..m. a.

6..*

b. e. i r a. ". a. s..C i.S

'l s..

.....;..n

.....i.

. s.e u.'.s..,.4....

a.r.n.......

3.4.

4

.i.a

. 4 *

4.. p. s.

.a

.>.s.4..t

s

....j

..s....

a......o

..s

.,3.t 4.

.t

4..

. < 2 2

4..s i. i. v.

a

. (f.y... ~...

.4.....

3. 3.j o..s.. )

,..p y...i

'. y.a...e

. a..r.s.s 1 x fj

/ f j ',.

a e **

..a

.a

(

...4 a

s 9

'a,..a.s.a. a s *t.-...s.

3 y.s.da..,y

.a #.

~ 4. J. J... e... w. s

. a.a..

b. s # a. *s *. a.4..a. a,

s b i a..'

J as 4 7.

,.m.,

...o_,

With Feed Without Feed and Sleed and 31eed A

cm 9 x 10 6 9 x 10-4

'TML (first cerel 1 x 10-7 1 x 10-5 TML (mature) 1.3 x 10'!

1.4x10-i 1.2 x 10~*

LOSF Case 1*

1.8 x 10-3 LCSP Case 2*

1.5 x 10'#

1.509 x 10-4 S0 2

_

Case 1 - ene of the diesel generaters cannot energi:e a meter driven AFW train Case 2 - both diesel generaters can energi:e a mo:cr driven AFW train f'or The ecenemic incentives :an he calculated by taking the ex;csure time The ecencmic the first cere as ene year and fer mature operatien as ten years.

incentive is essentially the redu:t 4 n in the present worth (at startup) of They are shewn en the folicwing prcjected menitary lesses due to accidents.

diagram:

Case 1 513.aM Case 2 l

s no F13 no ?&B 510.7M 523.2M V

Case 1

$560.000' Case 2 S15M _g Imcreve HFI V

lelia::ility s

F3B F33

-M

~.. _

~

-.. ~

-14 Start with a CE plant that has This diagrim :an be underst:cd as follcws.

l generater that can no feed and bleed capability and enly cne diese It wculd be worth up to su;;cr: a :ct:r. driven auxiliary feedwater pump.

hat is new the non-

$13.aM to ena:le :ne sec:nd diesel genera: r :: power w add feed and bleed

$22.3M ::

It would be wcrth up *:

The safe:y c.rade A 1 pu=o.

he discussed.

The final "fix' has yet *:

capabil'. y, anc se f rth.

n er :;erational changes such that i

value was arrived at by postulating ues gd ced fr:m 1.5x10-'/yr ::

1.Cx10~*/yr.

the likeliheed of an S D core mel is re u liability Of MF! substantially, g

This sign: te achieved by either in:reving the reLOCA substantia reducing the frequency of very small s...q 4..

w..s. c., a.t i..4

..uis A..

2. w-5. 4..v.s. b.y <. s...

4.. - s.a..,.. t..y 4

w

.3 s

.. a. e... r '..= - s. #. a..y

..s.

a.e.

d 't *v a. r

."..a.

e. 4 n,.3., i..., u..r. -..3.:. y.rv.

h.-.

"..*.2

, 4,

i.

a::endant risks r

- r.ny i s.. y

'ie have already examined t e h

w.

r valve se:: in:) Or s: e Of each.

agi~- s

".a. s....a.".s

. s.: n... -..... 'a s.

e

..q.,.

..u3. w,

'.. ;..- q... j ]

c,. r. ay.....2.

. w 3...4. g,.,j e i g.

.... x.. r..

$., J..

4..

... s

~. w.... f :.. e..

A.

4

.. -,...-...3......,.g.

a.. w.. < e..

a

e.., <.ca< a. s,..e... S... s. e. t. w.,.. >.g 1

.gn t

<s..a,<4-w

. wi.

,..,, n...... :...,..,.n.y.....c.

s...

4 s.., n.... y 4.s

c. <....-..... t'.nss

? CRY.L*CA's

.n

..,..s..........,,,..un

.f iscia:in3 e...

.... cssibil':v.

.w.

a :c.:..y,

..s 4 s.... 3.,.. s..y 4......4.....

4..w..w.

. e. < v.,.,,..

4

(::de safety :r 70RV in *he If :he MP* :an f:rca ::en a pressure reitef va:ve

....y...;

.,3...se a

..-.....y, c.

..m w."... 4.

. e. r cus

t.

'. '".s.a '. a ed

...s........,

/ ' ' ". =. a.'. X, s e...,v '..... I.'s '.'..'. 'J. a '.'6. "'X v '. ". *..'

y.. a......,,

4-

r. o..d.

s. e.u.. t..w.o-% *c*..d.#-**

'.'. C.'.

~~...".'d.

.. a.

. *.==".

............r....s 1..]

.w

. I

.y.

r. g.=.

9

-e. ;

m.

1 15 pressurizer valve opens,the pressuri:er quench tsnk rupture disk blows, and If the valve sticks open (and cannet be isclated),

a small spill eccurs.

U the operaters must restart HPI. Spuricus HPI actuations are quite common.

We assume here that the frequency of spuricus HPI actuaticns which remain en 1cng enough to challenge a pressuri:er valve is one per year.

I Sorrowing from the prior analyses we can draw the fellcwing event trees for the high head HPI design:

Withcut PCRV (or PORY ieft blocked)

Safety Valve Closes HPI Restart Upon HPI Shutoff

} small spill at 1./yr Spurious HPI Actuation ilarge spill at 10 3/yr 1./yr 10,3 10-3

' core melt at 10~0/yr With ?CRV installed and unblocked PCRV C1cses Upen 31cck Vaive HPI Shutoff C1cses HPI Restart

, s=all spill at 1/yr 5:uricus MPI NCtuatiC" s=ali spill at 10-2/yr i./yr 10-2

'large scill at 10~ /yr 10-2

.. _ 3._.

10~. core zeit at 10~7/vr i

9

.is.-

t 4

Note tha: if a PWR has a PCRV and high head HPI, it is better to run with I.

the bicek valve cpen, so the isolatable PCRV can take the brunt of spuricus

.'; *..= a l a u "....' *.

w : +

4.., s = s w e l l a s '. a.a.d.w = *.*. r..=.a.s '. =.n.. '..d".c ad. '..CA ' s.

i h the c:re eit sequences caused by s;uricus MF! actua:f on in plants with h g head HF* is ac:*pta:ly small and can be made scalier still if the ?CR7 nly Iif ts (bi:ck valve left : pen).

It is rcughly balanC2d by : parable risk

..e.s. nc...en

. a..........ca..

w..

.u..se >was.9.s,.u.e PC%/

a

.y

.. s.,.. z... s. n. u. a. c. a.

s.

s.,...

..s u.... a..

H wever, we sh:u*d n::a that there is 'a real e :ncaic incentive :: avoid the d *.".. e a '...= n d.' a.*. e. = 1 syills.

s.w n.,. 3.,.., <.. r,....n c... =.J. k.. ~, *.".. *.

.d i s k.= n

.....,4.1i. n..li..=rs.=.".=.v#.-.

e. mail sr 'i.-.=. d.

i 3

d e

a,

,a e., s s..... a. s. o.,. s..y

- -. e..= r...c.

~ #. a.x, e r...= d. 1..e=.s r.

. a. r.= '. a.. =. - r '# l,.'. e n '.". a.

i a

n.o 2. y u...-a i

.... r.... s.. -..,...... a.. n >i n.. e s,. 2. s <.,. s :.,..

u r

a..

a...

S/even x 10 year ex;;:ure = 350 milli:n fr:m the 1 even-/yr x 5x10 l

.3 4.

3.j.w.,.s..e.

..,. 4....

....e 4

i 4..

e.........

e.,.,

.. c.. s.

4.s..,.,..... *)....av..

w n..

( 510-s i

Wi..cu- :CEV: 10~~/vr i x 10: $ <.even: x 1.-.. vr =

~

510 With PCRY:

10/yr

(

',-***=.a.

.4

..5

'.#.....#**'.=.d.

..s.a.4.:.$.' =.s

. -a. s ". '.' s... '...=s#..-..'.'.'..=...

... - a.r.. '. v a.

t...

2.-'.

.......e.

h. a...,...a. c.

j.'*. 2 -.. e e ".

4..-. e a

g 3=.,e 3 * :.* m.. '. *s '. #...a. i.'.".

.3....=.. y

.r a. a

...vg..J..........J.

a.... i...,..

'J..* ?.......J....b...*.3.

,J....

2.=......j g /.

... i. s. #. 7g.j.

a.

1 9

17

-. (.

There appears to be no econcaic penalty (o,ther than first cost) in providing HPI pumps whose shutoff head is at normal RCS pressure, i.e., around 2250 psi.

1 In su=ary, then, this limited risk analysis cannot distinguish a dif.ference instali in safety among the several ways to achieve f aed and bleed capability:

one or more large PCRV's, raise the HPI head aboye the pressu izer safety valve setpoint, or install a smaller. FORY and raise the HPI head tr near These choices must be made on the basis of design r.ormal operating pressures.

adequacy or thermal hydraulic consider *.ciens, preferably considering ATWS as well as the design to assure that very. small LOCA's can be mitigated even though HPI or AFW may be late in starting or might be throttled temporarily by We have, however, found a plant availability incentive to the operators.

No -

avoid an HPI head so high that it can lift a pressuri:er relief valve.

such penalty accrues co HPI designs with a shutoff head at the normal RCS pressure.

I l

l F

[

a m_=__-

l

)

ATTACHMENT 4 UNITED STATES amseco

/,

'.o, NUCt. EAR REGut.ATORY CCMMiss!CN wAssincres, o. c. :csss y h s.,, 3 g

...%.,5'dj!

/c. c shQS' g.s.

January 22, 1982 v

MEMCPJNCUM FCR:

&, Eisenhut, CL, NRR

5. Hanaver, OST, NRR R. Mattsen, OSI, NRR C. Michelsen, AECD T. Murley, RCGR H. Thcmpsen, OHFS, NRR R. Vc11:er, DE, NRR FRCM:

Rchert M. Bernerc, Directer Division of Risk analysis Office cf Nuclear Regulatory Research ACCICENT SEQUENCE FREC'JRSCR FRCG?>1i CRAFT llE?CRT SU3 JECT:

The attached Accident Secuence Precurscr repcrt is currently being edited}

by ORNL with expected publicaticn in late March 1982. '4e are previding a The limited distributien of the draft rescrt for information purgeses.

techniques and methodology used in the report are sc ewhat con roversial.

For examcle, a questien has been raised of whether the c:rrect pr:babilities (absciute vs. conciticnal) were calculated and used to determine severe core mel: pr:bability. 'Je are reviewing this and other metheccicgy questicns within CPA. The Precurscr Report tends to indicate a ccre melt probability The repcrt i i

it higher than calculated in typicgi Feis.pr:babilityintherangecf10-/ reac

/ react:r/ year for typical F Js. The precurscr program tentative findings were presanted by CRNL (Jce Minarek) to NRC in meetings en 9/18/81 and 12/9/31. Two earlier draft versiens cf this repcrt wers given limited distributien within NRC, the first in early 1981 and the sec:nd draft reper: was distributed fc110 wing the 12/9/31 reeting. '4e have indicated to CRNL that we will previde them timely c:: men s befcre reper: publication. Please provide ss with any c::n ents ycu may have en this reper; by February 20, 1982.

,f/

/

w y/

/

7 Rcber: M. Bernere, Direct:r Divisien of Risk Analysis Office cf Nuclear Regulat:ry Research

Attachment:

As Stated cc:

R. Cennig, AECD 4

~

0. Ckrant, ACRS D.- 3 css, Rg ApRasang y L.

ng, ns 6L A. Thadani, RRA3, NRR 3

4 i

i DRAFT NURIG/CR-2497 Volt.:=e 1 CKNL/NS!C-132

\\

U i

3.

,.,3..ceRS 0

=N.

..A.9

. m: -

p-4

.M6 w b

A i.

CORI OAv. AGE ACO OI:CS:

1969-1979 Joseph 'J..1:arick Casi=1; A. Kukielka I

1 4

4

. :he U.S. :'celes: Eag=la:::7 Cc issi:=

7:e;:ared f::

Cf fice =f Nuclear le;:la:::7 lesear:h C=de ".::erare:07 Agree =e::s OCI aC-f!*.-73 and I.C-350 75 1

l m

m*t.r...u.A..at4 m... m. 7..

v.t.c e s=.L a

T

.a h.=w w==.* =t m

i A.

w s

.N

. 7

.se

.m,

- -p

e.

A3ST2AC"'

9 f

Descriptio:s of 170 :yers:ie:21 events, :syceted as I.I7.s, whi:1.ce-c ::sd at ce=sercisi lish: vste r e se c ;iss:s d::i s 1969-1979 a:d which to severs : ors ds=sgs s== ;;e-are ec sidersd to be potentisi ; setzscrs

's e nt ad, slo:g vi:1 associa:ed event ::ses and estageri:s:icus a:4 subse-q:s : ses17 s e s.

72.s repert s:==srizes ve:h is (1) de deve;c;=en: of

e dods. sad t o sc se: shes: 19,400 122 abs :se:s f:: po::::isi ;; cs:-

I scrs, (1) the isitisi s ar s e:ist of tho se abs t= set s to de :::=i=e whi=h 4

th :Id *:e reviewed is de:sil, (3) the de ssiled ::vist of. hose selected e

l 127.s which des yielded the 1~0 eve::s, ( 4 )' de : stag:ri:s:le: :f the 170 syst s, (fi the :al::1sti: :f f:::: ion f ail.: es ti=s ::s based c: ;;se:--

esti=stes to ser da:s, ( 6) the use of 7:cb sbility--of-severs sc s ds:s gs n

ssk ;;ee :scr events s:4 de id ::ifiestics of events : : sider.d sig-1 zificast, (7) ?:::ds s:slyses of sig=ificas: ev e nts, (3) th e id:::ift:s-t,

1: of the oder sve::s of i= ::ss whid me::::ed vi di: c

h f si;:ifiess: eves::, s:d ( 3) es1:.istics of 2: e sti=s te of seve rs :::e ds_-

age ;;:bability pe: r e ac t e: 7es: based c de sve:: : sshi g s.

l

L'E 70 ~2E '/CL'.'ME 0, 7. !5. E.:03.7, ;NLy Tug ;33 :,:-- : = r -.; :: ;-

,,.,n...,,___..a,,

. u...r:..,y

u.. :. :.:..:r.... e..,.,f r.....,.

s i

n m

n

ENCLOSURE SS

/ p* "*%**,

UNITED STATES g ger i

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. C555 W

.E

=

Cocket No. STN S0 470 FE!RUARY

! M2 Mr. A. E. Scherer, Director Nuclear Licensing C mbustien Engineering !nc.

1000 Pres;ect Hill Read Wi. dsor, Conn. 05095 Cear Mr. Scherer:

In its letter to the Chairman dated Cecember 15,1981, the ACRS ex ressed Tne NRC concern over certain aspects of the System 50 design (CESSAR).

staff is currently addressing these ACRS c =ents.

Specifically, we are addressing the c ncern that the System 80 design does not include ca ability for rapid direct depressurization of the primary system to allcw feed and bleed creratiens and the reliance laced caen the sec:ndary Tne Divisien of Systems Integration system for heat removal cacability.

(CSI) in the Office of Nuclear Reacter Regulatien (NRR) will be generating a Supplemen'.al Safety Evaluation Recert for Palo Verde and(Enclosure 1-).

CESSAR en this issue. Tneir current draft re:crt is attached.

Since CE has been reviewing the benefits of PCRV's fer the Syste'm 30 and since this issue is not c moletely resolved, we recuest that you review the attached infor atien and provide us ycur analysis of the need for PORV's in the System 50' design. Tnis infermaticn is being requested

-in sue:crt of the CESSAR review and will be the subject of future discussiens with the NRC staff.,

Sincerely,

(

\\

. Ei's k

ect:r Divisi:n of Licensing Office cf Nuclear React:r ReSuiatien Encl:sure:

1.

NRR Sucol. Safety Eval. Re ort.

NU$'

e

_-,___QO"*

WO_yO

4 P9 &

WG w

a e

se wg,,

,mgp g.

g 4

..,

ML20040D116

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