ML19206A325

Summary of 790411-12 Meeting w/C-E in Bethesda,Md Re short- Term Corrective Actions to Be Implemented as Result of TMI-2 Incident

ML19206A325
Person / Time
Site:	Crane
Issue date:	04/12/1979
From:	Ross D Office of Nuclear Reactor Regulation
To:	Case E Office of Nuclear Reactor Regulation
References
NUDOCS 7904190314
Download: ML19206A325 (16)

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- ",g APRIL 1 2 1979 MEMORANDU:1 FOR:

E. G. Case, Deputy Director Office of Nuclear Reactor Regulation FROM:

D. F. Ross, Deputy Director Division of Project "anagement SU3 JECT:

SUS' MARY OF MEETING WITH COMBUSTIC1 ENGINEERING (CE)-

CORRECTIVE ACTICNS FOR CC"BUSTION ENSINEERING NSSS PLANTS AS A RESULT OF THREE !!ILE ISLAND UNIT 2 INCIDENT On Aoril 11 and 12,1979, the NRC staff me* with respresentatives of Corbustion Engineering, Incorporated (LE) in Cethesda, Maryland, to discuss short tern corrective actions to be imple'"ented at CE pressurized water reactors (F', R) as a result of the incident at Three Mile Island Unit 2.

Several CE FWR licensees were in attendance. A list of attendees is attached (Enclosure 1).

April ll, 1979 "eetire The meetina ocened with an overview of the events at Three Mile Island Unit 2 (TMI-2) which require inrediate attention by all operating P',;Rs as these events are perceived by the staff in light af information available at this tire. These events are identilied as Items 1 thru 12 in the NRC Uffice of Inspection and Enforcorrot (01&E)Bulletin 79-02 of "pril E,1:P (Encic:ure 2). The staff speciticcil;. r.ctad that the responsibility f or developr.ent of corrective actions for these iters rests with CE and the utilities. The corrective actions that are needed are specific instructions to be issued i rediately to licensees of CE FW'Is.

These corrective r.casures will be reviewed by the NRC staff and issued by means of an OISE Bulletin.

D. Ross of the NRC staff read through the itens in the B&W bulletin (Enclosure 2) and asked for coments and agreement that the items were cuitable for CE designed plants.

Items 1 through 3 were agreed to by CE (agreement meaning that tne item was appropriate for a CE designed plant).

Item a censists cr four parts. They concern the overriding of autcratic actions of engineered safety features (ESF),

CE.es urailling te sceak on this iten tecause they ccnsidered this a prercgati've of their custorers. Tney stated that they provide guice-lines for tne cevelcpment cf prccedures by the utilities.

79041903N eJb u ccJ

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9 APRIL 1 2 1373 E. G. Case

-c CE stated that SI actuation occurs on the following signals (among others).

1.

low primary pressurizer pressure 2.

high containment pressure Containment isolation is actuated by the high containment pressure (with the same set point as for SI actuatica).

CE ' stated that while pressurizer level is the nain paraceter locked at in the guidelines, other system parareters are also used by some CE custoners.

CE has not provided a guiceiine for turning off the safety injection to its custoners.

CE agreed with item 43 (the operator shculd not override automatic actions of emergency safeguards fea tures),

A rcpr:scolative of Caltimore Gas & Electric Corpany questioned the reason for a time requirement in iten 4b cencerning conditions under which the HPI could be turned of#.

He cuestioned why a pressure indication was not sufficient. The poin,t was that if the Reactor Coolant System (RCS) were to approach guing solid, pressure (subccolirg) indicaticq would be sufficient, regardless of the time that the HPI had been in operation. Tne staff noted his comment.

The ouestion of the HPI causing reactor vessel pressure in excess of the allowable limits was discussed.

CE stated that they had perforned some fracture mechan;c: calculations of this type for the steam line break for vessels of different ages and these shcwcd accepteble results.

The analyses were previously reported to the staff in the letters listed below.

1.

June 4,1975 letter from W. Corcoran, CE t) R. "a c c a ry, N R C.

2.

June 24, 1975 latter fron W. Corcoran, CE to F. Schrceder, NRC.

CE stated that reactors 'esigned by them have not experienced a stuck open relief valve.

For this event, the pressurizer level could increase, although the primary system could be depressurizing.

u,3.

4 C{j

E. G. Case APhlt 1 2 1573 CE related a case in shich the system drained dcwn to a pressurizer level of l~

due to a drain valve that was inadvertently lef t open during a test (pre-nuclea r). There was no flashing or change of direction in pressurizer level.

However, this event was not directly applicable to the discussicn since the drain was releasing water inventury from the water space rather than stean spaco.

CE stated that the analysis of the inadvertent opening of a relief valve showed that the two phase level reached the top of the pressurizer with a void fracticn in the pressurize of approxia-tely 25' CE stated that under this condition they would expect the pressurizer level instru entation to give a true indication of lesel.

The staff asked '.,hether the operator had to take any action based on l evel,

CE restcoded t",t he did not and furthermore, that level did not ente. into any safety syste actions.

The staff then asled CE if the pressurizer could be full.. nile the core was errty.

CE replied that the analysis of the inadvertcnt opening of a relief valve sho'. sed that the pressurizer would drain.

The staff cuestioned '.Jrether the cc puter ccde used for the analysis would have predicted t'ris phenomenon.

CE stated that the code c.as an (valuatico ocdel co;e. The staff will pursue this further aith CE.

This completed the discussicn of Iter 4 Itea 5 concerned, erifying th" the auxiliary feed'.sater ( AFW) val /cs are in the propE* positicn.

CE agreed with this item and state" further that the AF's systens on all CE operating plants are enually actue ted.

Analyses of cli transients in the FSER shcwed that with a delay of 10 minutes be fore initiating TJW ficw, the pressurizer relief valves would not open.

CE agt eed that items 6 thrcugh 10 of the DILE Euiletir were apprcpriate.

CE had stated earl':- toet they had not ccse prepared to answer these i tcma in detail.

They stated that the inswers they gave could be considered tLose of the Corpcraticn howe /er The staff held c caucus and decided that oce infor aticn was recuired in order to write the bulletin, and CE xas given an addit:cre' 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> to ake cc rnnts in the folic' sing areas:

c,-

E. G. Case APRIL 12 379 1.

ESF resr.t criteria 2.

What criteria are now used for turning off HPI? What tenperature and pressure criteria are used?

3.

Should NRC advise utilities ca CVCS operation?

4.

Further discussion of fracture mechanics.

5.

Criteria for tripping the reacter coolant pura.

6.

List operator actions based cn pressurizer level.

7.

Provide a CEFLASH calculation for a small break on the pressurizer fluic dynanics.

It was decided to continue the meeting cn A ril 12,1979, at 1 : 30 pc April 12,1979 eeting The meetina opened..ith a rcview of the 'RC intentions 'to issuc bulletins to Westinghouse Ele:tric Corporation and Cerbusticn Ergineering reactor licensees by torcerow, April 13, 1979.

It was noted that Bulletin 79-06 was issued cn April 11, 1979, and contained general guidance for all P..'R reactor licensees (except Babccck s ', ilcox plant licensees),

The bulletins to be issued temorrcw will be based en but provide more specific guidance then Bulletin 79-06.

The meeting then nrr caded to the seven point agenda idcntified at tho end of the previous day's reeting. These seven points corresponc to the provisions of itcm of IE Eulletin 79-05 A.

The CE representatives discussed these provisions as follo'..s.

4.a. (agenda iten 1) CE agreed that this was appropriate for their facilities and suggested additional clarifying ins'. ructions to the plant operator. The staff will consider tnis s'uggesticn-4.b. (1) and (2) (agenda items 2, 3 and 4) CE stated that they ha /e been investigating these provisions since the

I-2 incident, and they confirmed the information given to us in the Acril il ~eetirg-regarding fracture echanics.

Eased en t"ei r investigaticrs and the referenced inferraticn (identified in tre A:ril li, 1979 minutes), they agreec that the provisicns of iter ' were acarccria:e for their facilities.

r-,3 4 0 43 sy.

.. ~. ~

E. G. Case AFRll 1 2 1979 4.c. (agenda item 5) The CE representatives believe that additional clarification sh7uld be added to this provision such that not all reactor coolant pumps should be run, but at least one pump cer cooling loop should be run.

It was recognized that, for s:Te facilities, the proviso that one pump per loop be run would require all reactor coolant pumas to be run.

The staff will consider this addition.

4.d. (agenda item 6 & 7) CE agrees with this provision as it is presently worded.

Regarding their prergency care cooling code (CE FLASH) calculation for a small pressurizer break, CE representatives state that some inforcatian is available on the CE System S0 docket (Safety Analysis Report Chapter 6).

This information provides pressurizer pressure as a function of time.

CE will, after a check of proprietary considerations, make available infor;. coon regarding pr essurizer level as a functicr. of t h a for this small pressurizer steam space brea'< (equi'calent to a 4" dia, hole).

Based or, the infurmatic" presented by CE, the staf f intends to proceed with issuance of a bulletin with short tern correc.tive actions to CE facility licensees, m

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i Denwund F Ross, Ceputy Director Divisior, of Project Management

Enclosures:

As stated cc w/ encl:

See next pa c i.<t 4 C9

E. G. Case afrit i 2 1979 Distributien Docket (50-320) t;RC PDR Local PER DDR Reading f4RR Reading H. R. Denton V. Stello R Volimer W. Russell B. Grinc:

T. J. Ca rter D. G. Eisenhut A. Schwencer D. L. Zierann P. Check G. C. Laina D. K. Davis T. A. Ic;olito R. W. Reid V. Noonao G. Knighton M. Fletcher D. Erinkcan Atto rr,ey, OE'.?

R. Frale, ACRS(15)

J. R. E;:har.an TERA I4RC Participants

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_ LIST OF ATTENDEES CCMBUSTION EtiGItiEERING MEETING 0d/11/79 FC-Baltimore Gas & Electric F. Fletcher, DOR C. H. Cruse I'

Lobel, DOR R. C. L. Olson

r. Orr, DSS A. Ignatonis, DSS Corbustion Engineerina F. Combs, OCA S. Droggitis, CCA R. G. Walker G. Sauter, 001 L vlinn J. L. Cre',is, I&E Region V C'.n' ?r,1

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,i. C.,,.oseley, 1st

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J. Crawford S. Diab, DO W. E. Burchill u.koss,DP".,.j C. B. Brinkman

{.G. Case,.

R. S. Dalees thadani, Cn n.

g~ tgngg D. Crutch #ield, N R D.yillavia,CF

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Consumers Pcxer rcr,3any b.

! ssallo, tw.,

R. S. Boy d, Li'j..

D. J. Va nde'.,'a l l e D. Skovho t t, Dr ri. P. Hoffran L. P. Crocker, D0" F. Schroeder, DSS L P O-S. H Hanauer, DSS L. B. Engle, CPM J. Ccstello S. w a rga, C r.,

D.C.Dilanni,CO)5-Omaha Public Po'. ar Dis trict J. Szukewicz, 7 (Pickard, Lcne 7,

Ga r ri c k )'

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P. F. Lollins, DC '

J. J. Holman, DP:j T. R. Rcbbins R. D. Sil ver, La E. Mckerne NUS Corocratior P. S. Kaco,

.;0-Iendonca, CCR C.,.]ielson, von G. C. Millman R. L. Denning,

.2A Yankee Atomic Electric ( 'aine Yankee)

D. G. Eisenhut, CCR T. A. Iccolito, OCR W. H Reed R.

P-Snaicer, CDR r '

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""D' Babccck 5 'lilcc)

Paulson

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Northeas: tilities

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Enclosure la LIST OF ATTENDEES COMBUSTION ENGINEERING /NRC MEETING 04/12/79 NRC Baltimore Gas & Electric M. H. Fletcher, DDR R. E. Denton F. Orr, DSS C. H. Cruse F. Combs, OCA R. M. Douglass E. L. Conner, COR R. C. L. Olson E. McKenna, DOP S. Diab, D0R Bechtel J. F. S tol z, C P" L. Engle, DPM B. T. Ritter R. P. Snaider, COR P. Schwartz R. K. Major, ACRS G. J. Falibota CC-bustion Engineering Northeast Utilities W. E. Burchill P. L'Heureux J. D. Crawford D. A. Kreps C. B. drinkman J. V. Strinaitis R. S. Bell C. L. Kling V. M Callahar F. L. Carpentir.o D. Ayres R. G. Walker

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Preliminary infor.mation re:cived by the NRC since issuance of IE Eulietin 79-05 cn April 1,1979 has identified six potential humar.,

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Tripping of reactor coolant pumps during the course of the transient, g

to protect against pump damage due to pump vibrction, led to fuel damne since voids in the reactor coolant system prevented natural circulation.

=

Actions To Se Taken by Licensees:

For all Sabcc k and Wilccx pressurized water reactor facilities with an

.perating license (the actions specified below replace those specified 4

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• b. C ro ^ ^

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.. +.

C.,

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c..; o r 4. : *i l y.

C. + e.

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e. b. 4. } j ';s i.

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=."-.i.

2 i

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.-w

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such Voids are for~ed.

w.}.9 '

• 1. O..,:

at-ouise in /v-U:n April b, 19/9 Page 3 of 5 4.

(This item clarifies and expands upon item 4. of IE Eulletin 79-05.)

Review the acticns directed by the operating procedures and training instru-tiens to ensure that:

Operators do not override automatic actions of engineered a.

sa fety features.

b.

Operating procedures currently, or are revised to, specify that if the hich pressure injecticn (HPI) syste-has been au cmatically actuated because cf Icw pressure condition,

.it must rer.ain in cperaticn until eitner:

(1)

E th ic'x pressure injection (LFI) pumps are in cperaticn 2a c.,i nin3 n

4

c. a rc.-.e in ex e.e.e

, n -.,

r - = c - s. c n,a

+ b.:-

w m.

situaticn has been stable fcr 20 minutes, cr 1

(c) as itL. p p..; i. 5 ;,. e. 3..

h' c. s L s

a Cg /

s. c...

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t.or

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d.

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• n.

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w w

r i

p.

i rs o i c.: '. i ^. d i o-c.v.=!".=

g-y l.=. '. c ^.r. w~ '. '. 4 m.,

e.. c., w.'. =. r d

i v; o

inventcry in the reactor primary system.

~ ;210.-

1.;..

- :. =.. =.

.r :

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.c, J 3 l e. + :s o- /.c.e:

i...,,

6 v.

)

Ve r a f,v "w o = ' e.. o. - r =. v,

1. . e =. " a. '. c. r. >. 'I v a. a-c. =. ir. '.." o.

a ' e..i - s#'.#c- 'io i

i or r.

e a o acccrdance with item 8 belce.

Alsc, review all sr.fety-related valve positicns an, csitioning rcc,uire cr.ts c assure t.

e nat valvos ar= "si*ui.r-='

(. r a. n ~.. c l o s =. d ) i r. a ~m =.... =.. '..

c. n s. "m* = '.'.c.

. r o.

prcper-operaticn cf engineered safety features.

Alsc reviea T 3. l c- +. n. w rv-aw,,,es, s.. L cs..-sc.

a

C r m.e. i n. c..r. c..3 3.;

. n.e :.,

i

.r-c o.

o v. i w

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a

w. i \\f ) I 4.i O.e.

.$ P.$

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• 8*

r 1 8 - **

- - '" 'a f. *i..r. a r.s m o f n, ] ] u n. 4. 2 n c. w-n e e." r.V ~o.*.. i n '.,'l a. #..' d.

r r

o~

4

t. 9 A 4 O.

"y

,IE Eulletin 79-05A April 5, 1979

Pae,

,n o ;..c

s. t 6.

Review the containment is:laticn initiation design and procedures, and prepare and implement all changes necessary to cause containment isciation of all lines whose isolation does not degrade core cooling capability up;n automatic initiation of safety injection.

7.

T

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Enclosure I to IE Culletin 79-05A rpril 5, 1e/

PRELIMINARY CG.". v'a t VL vJ Ioc.-,n-..

v.1 4.,-,c.

Cn,lh., /,/

J/dv ML L j r :.,1 j 8.

d U,o. I L C On- -

eu-u cSTCe.

TIME (Approximate)

EVENT 4

about 4 ""

Loss of Condensate Fu p (t = 0)

Less of Feedaater surbine Trip

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2 TIP.E EVENT t = 8 min.

Auxi?iary feedwater ficw is initiated by opening clcsed valves t = 8 min. 18 sec.

Steam Generater B pressure reached minimum t = 8 min. 21 sec.

Steam Generator A pressure starts to recover t = 11 min Pressuri:er level indication comes back en scale and decreases t = 11 - 12 r i r..

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a t = 3 hcurs RCS pressure increases te 2153 psi and electrcratic relief valve Oper.ed t = 3.23 hcurs RC drain tar.k pressure spike of 5 psic t - 3.8 hcurs RC drain tar.k pressure spike cf 11 psi -

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3-TIME EVENT t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operator cpens electromatic relief valve to depressurize RCS to attempt ir.itiation of RHR at 400 psi t = 8 - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> RCS pressure decreases to attut 500 psi Core flood Tanks partially discha ;e t =.10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 28 psig ccntainner.t pressure spike, ccr.tair. rent s *r *..=." s i n i *. i a *s e- " a.. '. s +. - -. c ". a '.. = r.E -'. c. = '.. c '.

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