ML20114C780

Submits Assessment of Licensee Participation in Severe Accident Mitigation Feature of Zion/Indian Point Study,Per Technology Exchange Meetings Held to Date.Related Info Encl

ML20114C780
Person / Time
Site:	05000000, Indian Point
Issue date:	05/28/1980
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	Peoples D COMMONWEALTH EDISON CO.
Shared Package
ML20105B053	List: ML20114C780
References
FOIA-84-243 NUDOCS 8501300512
Download: ML20114C780 (10)

Text

_..... _ _. _ _ _ _ _....

^

t&

I I

Docket Nos: 50-247 50-286 50-295 50-304 Mr. D. L. Peoples Director for Nuclear Licensing P.O. Box 767

Chicago, Illinois 60690 i.

,..e

~

Dear Mr. P.eoples:

JThis letter is to bring to your attention our assessment of the licensee's

" participation in the Zion / Indian Point severe accident mitigation feature study, as reflected in the technology-exchange meetings held to date.

As we indicated in our letter of January 18, 1980, there are several measures that may be effective in mitigating the consequences of a major reactor accident involving core melt. Such measures include filtered-vented containment systems (FVCSs), hydrogen control systems, core retention devices (CRDs), and dedicated or bunkered decay heat removal systems. Based on the first three technology

' exchange meetings held as arranged for in our letter of April 10, 1980, we are concerned that little meaningful conceptual design work on the aforementioned mitigation features is in progress by the associated utilities and/or their contractors. Essentially all of the work presented at the first three

technology exchange meetings has been by NRC contractors. This direction h' described in our letter of Aprilis incon' istent with the purpose, objectives, and g s

10, 1980.

Although the NRC staff and its contractors are actively and vigorously pursuing severe accident mitigation features, we also expect that the associated utilities involved will devote the necessary resources to perfonn meaningful conceptual design work on all of the aforementioned systems. Only through our combined efforts will we be able to obtain a thorough and comprehensive understanding of the utility, advantages, disadvantages and practical problems j

'of such mitigation features.

We would like to reiterate NRC's policy regarding the Zion / Indian Point severe j

accident mitigation featur a study. This policy was emphasized by Dr. Meyer i

of the NRC staff in his opening conenents for the first technology-exchange l

meeting on May 7, 1980, and is as follows:

l l

"The operating mode of the licensing staff in reviewing a case is fn va tha licenenne in th44 gen, the 74*a 334 !aA437 Dg4qt ;t414t4,;,

to submit ti eir analysis and any other infonnation st oportive of t1eir i

"c556"and"in turn"fdt'"th#"dthff"td"F#iif#4"it'"shd"f#"cTNJU66tT6h"Wi tY"Eny' 4

a=*

..other..relata d..infomat. ton..to-inepen& ely... reach -its. conc.l eions-a 14.make-oavs >

g senc somu sus <,.74) nacx o24o

• ...==---'e=>-=a>a.

8501300S12 840809 PDR FOIA SHOLLYB4-243 PDR

~

l

;t c..g.,.

g. w r.:

. 3..

s.,%,, ymm+.

.='

..c.-

S *q%.~

s

}

e s

s u

p,. a gp'

.[-

~

D.1L.' Peoples'.

m.

its findings. It is.nomelly the applicant's responsibility to

-justify his. case by. being able to support any asstanptions used in

, s" Lhis analyses and conclusions.... A major ingredient in making

..;,; g technical juagments is.the technical product produced by the

,:v.f.,' licensee.".

if

~

?This is a t'raditionaI and. appropriate approach and any tendency to reverse this s

. by having NRC produce the product and the licensee's' audit -it is totally

.1

, 'linappropria,te. ~ p n

, ' ~ ~

s

, K, +" : c c;8ased.on ' discussion you had with NRC staff after the May 20, 1980 meeting. I

- %.T bunderstand.that you'~1ntend a redirection to a more active participation '

3 Jthrough submittal of a ' detailed mitigation-features reports by June 3,1980 O

?and by."conceptualfdesign" presentatior.s at the-next (June 3,1980) technology.

'l' exchange' meeting.$1hopeJthat'thesecontributionsare' substantive.

~

b

^

u '.'$Please'W @ise'$3 Y kh.6C=.%

yv i?

Ah4 l

us;within-two wNeks-of the mceipt of.this letter of your

.o c

adv v

, ~. future' planiof: action'in: regard to perfoming specific conceptual design

i Mwork for the7 Zion / Indian Point severe accident mitigation feature study.

~

43 3

.L f,Q ', '

Sincerely, s

w u.

w - n-q

,: c e,c

~,

_.A y.C. _.2 n.r s,

,. ;3c;~ g.

p.

m.. w a g.5;.n w n q % g,,cf.p. c, s;.. '. barrell G. Eisenhut, Director qys.

y. p;q y_c

.q, :

2 e

~ 7mc n.f w

... j.: n j_

m Division of Licensing Gi,f4.T.y?vT:. W M.MP3. '; Office of. Nuclear Reactor Regulation

r.. J

.S, n >

s.

.r n

~c w-+

m

^:

Yr W;14y gg7.} gglg.- )j u

y

. _}

',s

'rg, b

s,

y y

. Q, y 3'g g g.4 j'H':DentonJ iSalvo

'C. Kelber Docket Files c.

D

. (cc) Esse attached list' S. Acharaya LOCAL PDR l

4

[

/ <,, W "E? Case R. Shterry NRC POR

^^

1>

.'R. Mattson

'J. Curry

_ TERA q

o p ;D..Eisenhut

-F.

Rowsome

,.. E.DISTRI9tf710N;M;,, %

l0.'R6ss M " % y 3,

~

NSIC'

'N I

i

' iS.-Hanauer.

R. Bernero NRR RDG FILE

. ~1T.<Speis" s

>"W Russell W. Houston RSB RDG File-4, % C

"" Y IG. Zech;

.4J."Meyer.

W.'Gammill JJ."Long'

' ~

m

CJ. 01shinski' J. Butts

<A. Marchese

J. Carter.

, T. Murley Quittschreiber (6) ACRS

..E. - Fenstermacher 1R. Budnitz DL

-L. Soffer

~

A.'Schwencer DEisenhut W. Butler L. 01shan 5/ /80 E. Reeves M. Sijverber, & s

.,,,,,,y'Nb....f,,n,...,

,,0$

g, Q H,,,,,,,,,,,,,,,,,, g,,,,,,,,,,,,,,,,,

1.-

. DS I ;. M B,,,,,,,,,,,,

I f

i

'P k

Dross ~

J01shinski

==. * (

s,.

• 0..........

. ! n

.Q......... 5RhlBG.........Sf

/.80......... 5f....JBa....... Sl....]Ba.....

l _ ~ ~_..

1 1,'

e

~

Docket Mos: 50-247 50-286 50-295 50-304 l'

Mr. D. L. Peoples i

Director for Nuclear Licensing P.O. Box 767 I

. Chicago, Illinois 60690 Dear nr. peopies-L This letter is to bring to your attention our assessment of the licensee's i

participation in the Zion / Indian Point severe accident mitigation feature study, as reflected. in the technology-exchange meetings held to date.

As we indicated in~ our letter of January 18, 1980, there are several measures that may be effective in mitigating the consequences of a major reactor accident involving com melt. Such measures include filtered-vented containment systems t-(FVCSs),hydrogencontrolsystems,coreretentiondevices(CRDs),anddedicated or bunkered decay heat removal systems. Based on the first three technology exchange meetings held as' arranged for in our letter of April 10,1980, we are concerned that little meaningful conceptual design work on the aforementioned

~

. mitigation featums is in progress by the associated utilities and/or their contractors. Essentially all of the work presented at the first three technology exchange meetings has' been by NRC contractors. This direction

)

i-

'is 1

~ istent with the purpose objectives and goals of the meetings e descri in our letter, of Apr11.10,1980.

Although the NRC staff and its contractors are actively and vigorously pursuing severe accident mitigatir.n features, we also expect that the associated utilities involved will. devote the necessary resources to perform meaningful

~

conceptual design work on all of the aforementioned systems.. Only thmugh our combined efforts will we be able to obtain a thorough and comprehensive understanding of the utility, advantages, disadvantages and practical problems of'such mitigation features.

We would like to reiterate NRC's policy regarding the Zion / Indian Point severe Laccident mitigation feature study. This policy was emphasized by Dr. Meyer of.the NRC staff in his opening coussents for the first technology-exchange meeting on May 7,1980, and is as follows:

"The operating mode of the licensing staff in reviewing a case is for the licensees, in this case, the Zion and Indian Point ut111ttes, to submit their analysis and any other information supportive of their case and in turn for the staff to review it and in conjunction with any other related information to independently reach its conclusions and make e-em-emm--w r i',s--e

=es--arena mm

-e. e sr e -m e-um-na+-w--uman am oe.-

-awmeem--o'm==.sw--e=.wwweave s s--

t v se t n=e.sMi ww e erk evm-* wrw---W ww owtwwww*-=W-F--w--s**we'T-'

, ~

D. L. Peoples Its findings. It is normally the applicant's responsibility to justify his case by being able to support any assumptions used in A major ingredient in making his analyses and conclusions....

technical judgments is the technical product produced by the licensee."

This is a traditional and appropriate approach and any tendency to reverse this 2

by having' MRC produce the product and the licensee's audit it is totally inappropriate.

BasedJon ' discussion you had with NRC staff after the May 20.-1980 meeting I understand that you intend a redirection to a more active participation

through-submittal of a detailed mitigation-features reports by June 3,1980 and by " conceptual design" presentations at the next (June 3,1980) technolggy-I exchange meeting. I hope that these contributions am substantive.

l Please ' advise us'withN two Mks of the receipt of this letter of your future plan of action in regard to performing specific conceptual design work for the Zion / Indian Point severe accident mitigation' feature study.

Sincerely, Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation

• ma*

R. DiSalvo C. Kelber Docket Fil'es cc: see attached list H. Denton S. Acharaya LOCAL PDR E. Case R. Sherry NRC PDR DISTRIBUTION:

D. Eisenhut F. Rowsome TERA e

D. Ross-R. Mattson J. Curry NSIC g!

T. Spets S. Hanauer R. Bernero NRR RDG FILE J. Meyer~

W. Russell W. Houston GRB-RDG File J. Long

'G. Zech W. Gamill

$l A. Marchese J. Olshinski J. Butts J. Carter T. Murley Quittschreiber (6) dLCRS 1

E. Fenstermacher R. Budnitz DEisenhut L. Soffer A. Schwencer 5/ /80 W. Butler L. Olshan E. Reeves M. Silverberg DSI:RSB DSI:RSB DSI:RSB DSI:PS DSI

.DL AMarchese:mf 'JMeyer TSpels PCheck Dross 401shinski q

\\

5/ /80 5/ '/80 5/ /80 5/ /80 5/ /80 d/ /80 I

1 p

~~

~

Decket Mos: 50-247 50-286-50-295~

(

go. sos.

Mr. D. L Peop'les

= Dimeter for Nuclear Licensing I

,f P.O. Box 767

Chicagoi 1111asiso 60690.

o Af[,w r g %, ;. y %, f wx Deer Mr.sPeoples O, P

4 y

• W; ;

• ap t*. ' '.-

~

R-m L

This letter is to brin 6 to your attention our assessment of the licensee's

. participation in the Z< en/ Indian Point l severe accident mitigation feature

..stu#.-as reflected in the technology-exchange meetings held to date.

y>3,9 I

As we indicated in our letter of January 18, 1980, there are several measures l

' that may be effective in mitigating the consequences of a mejor reactor accident involving core molt. - Such measures include filtered-vented containment systems (FVC$s) hydrogen control systems, core retention devices (CRDs), and dedicated

'or bunkered decay heat removal systems. Based on the first three technology exchange meetings held as arranged for in our letter of April 10,1980, we

'are concerned that little meaningfk1 conceptual design work on the aforementioned mitigation features is in progress by the associated utilities and/or their contracters. Essentially all of the work presented at the first three a. technology exahange meetings has been by MRC contractors. This direction t

X~

.is inconsistent with the purpose, objectives, and goals of the meetings described in our, letter of, April 10. 1900..

,y

' Although the NRC staff and its' contractors are actively and vigorously pursuing severe accident mitigation features, we also expect that the associated

' utilities involved will devote the necessary resources to perfom meaningful conceptual design work on all of,the afommentioned systems. Only thmugh sur combined efforts will we be able to obtain a thorough and comprehensive 4

' understanding of the utility, advantages, disadvantages and practical problems

.of such mitigation features.

~

We would like to reiterate NRC's policy regarding the Zion / Indian Point severe accident mitigation feature stu@. This policy was emphasized by Dr. Meyer of the NRC staff in his opening comments for the first technology-exchange meeting on May 7, 1980, and is as follows:

"The operating mode of the licensing staff in reviewing a case is for the licensees in this case, the Zion and Indian Point utilities, to submit their analysis and any other information supportive of their case and in turn for the staff to review it and in conjunction with any other related information to independently mach its conclusions and make 7

n

}

i i

.D..(Peoples '

i

.its findings. It 'is normally the applicant's responsibility to

~

. justify his case by being able to support any assumptions used in his analyses and conclusions.... A major ingredient in making technical judgments is.the technical product produced by the j

'11oensee." ~

~ $ ;

This is a traditional and appropriate approach and any tendency to reverse this "

by having NRC preshece the product and the licensee's audit it is totally

~y inappropriate.

gf '

Based os.disEussten you had with NRC staff after the May 20, 1980 meeting. I

- m understand:that you intend a redirection to a more active participation

,, through submittel-of a detailed mitigation-featutes reports by June 3.1980

and by.'senseptual design

• Presentations at the next (June 3,1980) technology-H,4, eschenge,meettag.Hl. hope that these contributions are substantive.

l

s.

<g MQtS4Wre @,

.y s

'Please advise.us:withinitue seeks of the receipt of this letter of your fkture plan of action la regard to performing specific conceptual design

- ^-.

uork fbr the Zion / Indian Point severe accident mitigation feature study.

Sincerely.

1

~

a,.

m La ym x ' d,,.sm '

,e v d 4

j

,g

- cm

-Darrell 8. Eisenhut. Director c.

Division of Licensing Jj

.3

-, '*/;; - gi

- e : Office of Nuclear Reactor Regulation 5 h k j l, W k h ;

b 2'

.y.3 s.wl % 't :. n q,:>,pr7'h 0; 4

lv3

• z _f.' ' h f. & ; ;-

T 3s

,v 9_7,;

g r.

', - - cc. see attached list-R. DiSalvo C. Kolbe.-

Pocket Files H. Denton

.S. Acharaya LOCAL PDR in 1

'v'

' ~ C E. Case'

'R. Sherry NRC PDR DISTRIBUTION:'

~

. D. Eisenhut F. Rowsome TERA 1

D. Ross R._Mattson J. Curry NSIC

.S. Hanauer' R. Bernero NRR RDG FILE cT.:Speis r

J. Meyer

~

W. Russell W. Houston GRB-ROG File J. Long

.G. Zech W. Gannill A. Marchese

'J. 01shinski J. Butts J. Carter T. Murley Quittschreiber (6) 4tRS R. Budnitz DEisenhut E. Fenstennacher L. Soffer A. Schwencer 5/ /80 W. ' Butler L. Olshan E. Reeves M. Silverberg DSI:RSB DSI:RSB DSI:RS8 DSI:PS DSI DL AMarchese:af JMeyer TSpeis PCheck Dross J01shinski 5/ /80 5/ /80 5/ /80 5/ /80 5/ /80 5/ /80 n.,

---

n~ -

- - - -. - -,n.---,,,~+,--,---n.-,w.~--,n,,--,aa-,

o-

.m,n,_

-w

~

BROOKHAVEN NATIONAL LABORATORY MEMORANDUM DATE:

April 27, 1981 To:

. Holders of BNL-NUREG-28750 mou:

A. J. Bus 11k and R. A. Bari ej

/-

sus;EcT:

Erratus This memo is an errate for Reference 1.

An error had been made in abstracting the maintenance unavailability for a High Pressure Injec-l tion System pop from the Reactor Safety Study. The correct value of qg is.019, not.057, as was given in Reference 1.

There were, in addition, typographical errors in Eq. (17), and in a ntaber given in the ninth line I

of the second paragraph on p. 35 of Reference 1.

This last mentioned num-ber should be 1.3x10-3, not 1.3x10-2 The error in g propagates through the report.

Equation (26) has

}

g 1.9x10-2, not 5.7x10-2, on the right-hand side. Equation (33) has 8x10-4, not 1.6x10-3, on the right-hand side.

Equation (38) has 2.5x10-3, not 4.8x10-3, on the right-hand side.

The probabilities for accident sequences S D and S D are changed, which results in changes to pages 37 and 38, to I

2 Table 3 on p. 43, and to the listing of contributions to the release cate-gory probabilities of small loss of coolant accidents given on p. 47. The f-

_ corrected pages 37, 38, 43, and 47 are enclosed; in addition, a corrected L

page 32 is enclosed, which corrects the typographical errors in Eq. (17).

Reference l

1.

A.J. Bus 11k and R.A. Bari, "A Critique of the Offshore Power Systems Risk Study for the Zion Nuclear Power Plant", BNL-NUREG-28750, December 1980.

AJB:RAB/nn Attachments

./

kl c_

t

~

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

Y

2.3.2.2 Quantitative Analysis of the HPIS If we neglect the possibility of common cause failure of the HPIS pumps, then pr {G) = (qA + 4 I I9D + 9E I + 2 (9A + 9CI (93F + 9MI C

2 2

+ 2 (qD + 4 I I41F + M) + 2qip qM + 2q3F 9M (17I E

2 2

+ 8qip q3F g + 2qip q3F + 2qip q3F l

where qA " Pr {A}

(18) qC " Pr (C)

(19) qD = pr (D)

(20)

I qE = pr {E l

.(21) q1F = pr {Bf} = pr {B )

(22) 2 f

q3F = pr {B ) " Pr (Bb (23) 3 4

qM " Pr {B } = pr (B.1

• Pr (B } = pr (B }

(24) 2 I

{

In deriving Eq. (17) from Eq. (16), the unavailability of a HHSIP due to i

maintenance was assumed equal to that of a CCP.

)

Using the Reactor Safety Study failure data given on p. 300 of Appendix II of Reference 2, L

pr { pump FTS) = 2.5x10-2 (25) l where FTS means " fails to start", and

-2 q = pr {1 pump is out for maintenance) = 1.9x10 (26) g From Eq. (2),

I

= pr {Bf} = pr { lCCP #1 FTSI) tpr { ] }

q ip j 9

,-rm,,-

,,w--eemr,

-,v--,..,,---y an

+n,+

r,m.,,-ee-e-e.

e,--,-

~m-w-

we

r' l

Segur:nce S20 pr (HPIS Failed}

2.5 x10-3 1x10-3 reactor-yr pr { S )

/

2 pr (S D}

2.5x10-6/ reactor-yr 2

Sequences 2HFY pr (HPRS and CSRS Failed) 5.4x10-3 pr {0perator Error During Injection-Recire. Shift) 3x10-4 pr { }F}

5.7x10-3 pr {Y}

.9 1x10-3 reactor-yr

/

pr { S )

2 pr { 5 HFX) 5.1x10~0/ reactor-yr 2

Sequence SpHFX

-3 pr (HPRS and CSRS Failed) 5.4x10 pr (Operator Error During

_4 Injection-Recire. Shift) 3x10 pr (HF) 5.7x10-3 pr {X}

.1

-3 pr {S )

1x10 / reactor-yr 2

pr {S HFX}

5.7x10-7/ reactor-yr 2

The probabilities of the accident sequences S H and S H are taken to be the 1

2 same as given in the OPS study (Reference 2),

pr' (S H) = 1.2x10-6 reactor-yr 1

/

pr {S H1 = 3.9x10-6 reactor-yr 2

/

4 _.. _. _ _ _ _ _.

- J

3.0 DISCUSSION AND SU MARY The contributions of the loss-of-offsite-power initiated accident se-quences to the various release categories are given in Table 1.

The sequence TM'B B', involving failure of the reactor coolant pump seals, contributed a 5

probability of 2x10-6 reactor-yr to release category 2, while the accident

/

. sequence TMLBB'Bo, involving loss of auxiliary feedwater, contributed a probability of 6x10-7/ reactor-yr. There was a probability contribution of 2x10-6 reactor-yr to release category 5 from loss-of-offsite-power initiated

/

1

(

accident sequences. One should keep in mind that accidents where the contain-l

. ment failed by overpressure, but the containment sprays were operating, were assigned to release category 5, but there may well be a considerably larger radioactive release from accidents where the containment fails by overpressure

, before reactor vessel failure, as compared to accidents with containment fail-ure some time after reactor vessel failure (when the sprays have had time to l

remove the fission products from the containment atmosphere).

In this connec-i tion, failures of containment by hydrogen burning before reactor vessel melt-through are of special interest.

Such detailed analyses of containment failure mode probabilities and release category assignments are beyond the scope of the present report. The contributions to the release category probabilities of the loss of coolant accidents are, from Tables 3 and 4 1.6x10-6 reactor-yr l

Release Category 2:

/

Release Category 5:

9x10-7/ reactor-yr 6.7x10-6 reactor-yr Release Category 6:

/

Release Category 7:

8x10~6/ reactor-yr.

The loss of main feedwater initiator (see Section 2.2.4) gives a Category 5 re-lease of 1x10-7/ reactor-yr, while the Category 7 release is 9x10-7/ reactor-yr...

.