ML19325F169

Amend 72 to License DPR-22,revising Reactor Vessel Pressure Vs Temp Limit Curves to Meet Staff Positions of Reg Guide 1.99,Rev 2

ML19325F169
Person / Time
Site:	Monticello
Issue date:	11/02/1989
From:	Thoma J Office of Nuclear Reactor Regulation
To:
Shared Package
ML19325F170	List: ML19325F169 ML19327B995 ML19327B996
References
RTR-REGGD-01.099, RTR-REGGD-1.099 NUDOCS 8911140357
Download: ML19325F169 (13)

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I Plymouth Nuclear Matters Committee 1

Town of Plymcuth 11 Lincoln Street t

Plymouth, MA 02260 L

September 00, 1989 l

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Mr. Thomas E.

Murley

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Director buaw Office of Nuclear Reactor Regulation a v T (11 /

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df" Nuclear Re$ulatory Commission f0f@D 7920 Norfo k Avenue Bethesda, MD 20B14 HE:

P11 grin Nuclear Power Station Direct Torus Vent System J

Dear Mr. Murley,

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Please find enclosed, copies of correspondence reisting to the recently installed hardened wetwell vent at P11 gram Station.

i Since several of the issues under discussion concern NRC's review and approval of the system and since you are one i

of the key individuals involved in the modification,It would we are i

seeking your input to help clarify this situation.

be greatly appreciated if you could respond directly to relevant aspects of this issue in writing to the above address.

'The staff iound the Althou$edsystemandtheassociatedh Generic Letter 89-16 statesbECo analysis acceptable,'

instal we have not been able to conclude this from any of the other i

existing documentation.

Specifically, all of the Safety t

Evaluations describe only the installation, not the use of the in Safety Evaluation 2269 dated vent.

Also, the logic used 1/9/88, which concludes that a change to the Technical is very questionable.

Do you Specifications is not required concur with BECo's arguement tbere?

In addition, inadvertant or premature venting is a very serious safety question, yet, in various documentation, BECo maintains that the DTVS does not involve an unreviewed safety question.

If you agree, could ycu explain why it does not't Many state and local public officials, as vell as numerous residents realize the close and necessary linkage between controlled venting and emergency prepareoness.

However, as you may well know the adequacy of emergency planning for Pilgrim is hotly d,ebated.

The topic is even under investigation by the NRC Inspector General's office.

Do you believe thut the DTVS should have been allowed to be made operational without adequate emergency prepareness by the community and the licensee?

Obviously, this is a far reaching technical and politically sensitive issue within the NRC.

In revievano the wJ of course, would have preferred that the documentation, NRC approach to tbis issue had been more straightforward:

if it was a

idea, behind it and insure that at was designed,goodinstalled,get and planned for properly, and if it was a bad idea, stop it from being implemented.

However, the

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L existing documentation indicates official heestancy; no one seemed to relish having their names, reputations, and careers closely tied to this plant modification.

We live downwind of P11 gram.

lt would reassure us if you could provide your assurances that BECo is up to the task of using this powerful i

new tool.

Pilgrim, as you are aware, has had a very troubled history:

Some of the largest fines longest shutdowns, most expensive capital repairs, highest b & M costs, and lowest SALPs of currently operating reactors.

Now, the first DTVS in the nation is installed here and we are extremely concerned.

Also included is our report on the April 12, 1989 spill in the RCIC system at Pilgrim.

There are many issues here which we feel vill be of considerable interest to you.

First, the AIT report contained errors.

Second, theexecutivesummer(hebody the conclusions from and cover letter did not reflect of the report or from the appendixes.

Third, it was an Fourtb,a topic <with systems loss of coolant accident interfacinkavebeenclosely which you involved.

we are requesting higher level NRC review of the issue, with special emphasis on the role of NRC in the event investigation and, more broadly. In the power ascension oversight.

These are serious assertions and serious requests.

Your commentu on both of these matters would be greatly appreciated.

Thank you,

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C David C.

Dixon Vice-Chairman, Plymouth Nuclear Matters Committee 40

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e-TOWN OF PLYMOUTH i

1I Linec,ln Street Plymouth, Massachusetts 02360 (6171747 1620 t

september 5, 1989 j

Mr. David F. Tarantino District Manager l

Nuclear Information Division j

448 State Road, suite 5 Plymouth Ma, 02360

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Dear Mr. Tarantino,

Thank you for the information on the Direct Torus vent j

system.

Regretably, we already had obtained those i

documents, with the exception of the most recent letters i

between Peter Agnes and Ralph Bird, and the questions we had asked resulted from the study of those documents.

We now resubmit the questions and ask you to seek ditect responses to them.

I The significance of this issue should not be underestimated.

j Prior to the DTVS, one of the final layers of defense in j

depth was the steel and concrete Mark 1 containment, which l

has a burst pressure of over 100 psi. The DTVs punches j

through that layer, relieving directly to the environment at only 30 psi.

It is the most significant change to Mark 1 l

containment design in twenty years, and is the first such system in the nation. It use requires early notification and i

coordination with Civil Defense officials in the EPZ.

I While we would like specific responses to the ten questions, l

the most important isaues can be distilled into two main 1

creas i

1.

The NRC has indicated in neveral instancas that they were unwilling to endorse Pilgrim's DTVS and that the installation of valve A0-5025 would require a change to the Technical specifications.

In all of the documentation available to us, the installation and the use of the DTVs were analysad seperately.

Futher, BEco states repeatedly 1

that the system will not be made operational, that the valve A0-5025 will not be installed without formal NRC approval.

i The valve is now installed and operational.

Can you provide l

this committee specific documentation indicating that NRC has now formally approved the use of the Pilgrim DTys, that i

its use does not introduce unreviewed sataty questions and that BECo, in proceeding with the installation, has not violated 10 CFR 507

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2.

The logic behind using the DTVS is complicated.Yet, our reading of the docutaents indicates that there have been no changes to your EOP's incorporating the new decision trees or early notification requirements; no training on the use of the special keys, electrical' jumpers, special fuses; or the other idiosyncracies of the system.

No management review, no public involvement. Only the pre-existing EOP-3 relates to containment venting, and BEco did not rewrite it before implementing the new DTVS.

It detailed procedures

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have been prepared, please issue us a copy.

If not, please 1

explain why it is not necessary to prepare to use this powerful and potentially dangerous system.

j While we commend BEco for going beyond the NRC requirements for aitigating severe accidents beyond the design basis, we require assurances that the system has been implemented properly and that both the utility and the state and local groups are prepared for its use.

We have not obtained that assurance from the available documentation.

If you require clarification of this request, please write to our committee, care of the Town of Plymouth, or call committee member David Dixon at 508-946-1000 during the day.

Thank you, Plymouth Nuclear Matters Committee CC:

Ralph Bird, Sr. VP-Nuclear, BEco Plymouth selectmen l

Thomas Murley, NRC-NRR l

William Russell, NRC Region 1 Richard Wessaan, PDI-3/NRR Dan Mcdonald, NRC-NRR Charlie Marshall, Pilgrim Resident Inspector l

Members of the Nuclear safety and Health Advisory Committee l

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i To Plymouth Selectmen and Plymouth Nuclear Matters Committee Members I

Troms, David C.

Dixon

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Subject:

Request for Information on the PHPS Direct i

Torus Venting System Date:

June 13, 1989 Durina our tour of Pilgrim last month Mr. David Tarantino offered to have technical cu,estions about the Direct Torus Venting System (DTVS) answered by the engineering staff.

In response, our committee has developed the attached last of questions.

They were revneved and approved by committee during the May 24, 5

1989 meeting.

These questions have arisen from our study of the DTVS.

It is an important assue which has received little public discussion, an part due to its technical l

nature.

This vent releases pressurt, and possibly fassion products, from the containment durano a severe accident directly into the atmosphere, thus bypassing the inherent safety offered by the steel and concrete r

protective containment structure.

In theory at is to i

De used only as a stopgap measure to keep the, from rupturing thereby avoiding a more containment serious, uncontrollable release of fission products to r

the environment.

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There are three main issues in the analysis (1) Under what accident scenarios is the DTVS intended to be used, given that for some accidents it helps, j

some it exacerbates and others it's irrelevant?

(2) Has SECO implemented the concept properly?

Has at minimized the risks of improper use of the vent, such as inadvertent or premature ventinc7 Are their people i

t r a a rse d to use sucn a powerful tool should it ever become necessary?

Is the public prepared to respond?

i (3) Has the NRC played its proper role an this modification?

Since the modification exists to i'

mitigate accidents beyond the desion basis, the NRC has taken a hands-off approach.

Also, if the NRC had maintained its anatial assertions that the DTVS required a change to the Technical Specifications, puk'ac hearings could have been necessary.

We are -Fequesting this inf orma*. ion f rom SECO to enable I

us to issue a more complete report analyzing the DTVS.

Answers to these questions will fill in some of the gaps.

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TOWN OF PLYMOUTH I1 Uncoln Stree Plymouth. Massachusetts 02360 16171747 1620 e

June 2 1969 Mr. David Tarantino Pilgrim Nucitar Power Stataon Mocky Hill Road Plymouth, MA 02360

Dear Mr. Tarantano,

Thank you for guading us on the informatave tour of the station last month.

The tame spent allowed the e. embers of our comnattee to better understund the operation of the facalaty.

During the tour, you offered to accept questions of a technacal natur e hbout the direct torus vent.

The commattee has several questions for whach we would likt answers before proceedang wath our revavv of the DTVS.

The members of the committee belaeve that the DTVS as a powerful and somewhat controversaal tool whach could help the plant operators mitagate the effects of 3 s e ve r e-accament.

We need to acquare a better understanding of the system to help us evaluate the benefits and rashs of thac

- ins t a ll a t a cin.

Your written responst would be greatly appreciated.

Ebould you need to dascuss thas request for information, pleast feel free to call or wrate to o.'ie of our cornattee members: Davad C.

Daxon. 135 Gunners Exchange. flymouth, MA.

Day phones 946-loOO, ext.2497.

Eve phont: 7474 0963.

Thank'you again for your help an thic matter.

If at appear s that that request macht takt longer than two weekc to fulfill, please let our comnattet know whs'n we nach*.

expect a response.

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Sancerely.

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dL.L onet.

4TTErt.5 06w,

cc: Plymouth Selectmen

'P1 mouth Nuclear Matters Committee Members tr>

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l DTV51 -- 6/12/69 Page 1 j

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r GUEFT3ON 1:

Certaan actions are required to open the outboard containment valve AO-bO2b.

Could you indicate where the tuse installataon occurs to enable power to the DC solenoidi Also, who has possessaon of the key for the c

remote manual swatch whach opens valve AO-bO2b?

l GUEST 3ON 2 Certaan actaenc are reoutred to open the inboard contaanment valve AO-bO42B after the automasse contaannent high pressure trap poant has been schaeved.

Could you descrabe the manual anstallatson procedure for the hard ware Jumptr7 Does this action occur behind the panel an the control room, or out an the plant?

9UESTION 3:

The earlier design for the DTVS also had an automatic reclosure of the vent if a high redsation level an the torus was achaeved (1).

Thas as now deleted from the current desagn (4).

Could you andacate why this safety element of the desagn was elamanated?

QUESTION 4:

The rupture disk in the vent line is specifatd for 30 pea ( 3 ).

Yet the contaanment design pressure as approxametely 60 psa and ultamate rupture pressure of the contaannent as approxamately 120 psi.

Could you explean why the DTVS as an*.endec to operate at such a lov rsure?

QUESTION 5:

Are thtre desion basis accadents for whach at as calculated that the torus pressure could exceed 30 psa?

OUEST20N 6:

In early correspcndence wath the NRC, BECO andacated that anformation on procedural changes assocasted with the modafacation for the DTVS would be physicalprovaded(plantLater correspondence ac calent on this 1).

matter.

Have procedures controlling the use of the DTVS been completed, revaeved and approved by BEC07 Have these procedurts been revaewed or sporoved by the NRC7 How many and who of the PNFS persennel have been trained and have fornally signed off en the procedures?

Can a copy of t h c-procedures be made available to our committee?

GUESTION 7:

During the March 7, 1988 tour of PNPS by Mr. Russ&11, Dr.

Murley, and Dr. Thadana, BECO responded to the cuestions gestd by the NRC in their 'Instial Assesspent of F11 gram In that presentation. PECO stressed that the

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declaration of central Emergency and recomnendations for protective actions wall be assued by BECO earJy an events l

which map lead to containment ventang(3).

Does PECO have I

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DTVS1 -- 6/12/89 Page 2

. approved guadelanes and procedures in effect for recommendang evacuataon of the EFZ an events which may lead to containment ventang?

Have the people in the who are charged to draft emergency actaon plans been,EP2 brief ed on tr.e DTVS and the a mpact of such early notafication and potential evocustaen?

GUESTION 4:

Mas any samilar venting system been installed and made operational at any other GE Mark 2.

22. Yankee proceedIII facalaty or in the U.S.

or elsewhere?

Dad. Vermont with a DTVGt Are there DTVS outsade the U.S.

which vent through carbon or gravel beds, resulting in a ground level releast on utalaty property?

Are there any DTVS operataonal which vent through a stack resulting in an seria2 dispersaon with potentaally grea,ter geogranhac contaminattent What are the pros and cone of eather arrangement?

DUESTION 9:

We request clarification of BECO's actaons in light of the NhC's stated positions on the DTVS.

In the NRC's anstaal assessment of the F11 gram Safety Enhancement Procram, the NRC was not prepared to endorse the use of the DTVS (2).

Further, the NRC stated that the anstalistson of an add 2tional branch line and conteanment asolation valve would require a chance to the plant Technscal Specafications(2).

Thuc the Nhc concluded that the anstallation of the DTVE could not be amplemented under the provas2ons of 10 CTR 50.59 (2).

However, the additionna branch line and the new outboard contmantent isolataen valve AO-5025 have been installed. SECO cleans that NRC approval is not required because, first, containment ventino has been previously approved an the Boalang Water Reactor Owners GroubO25 meets the NhcEmergency Operating Guide 12nes, and second, valve AO-requirements for a sealed closed asolatacn valve as defaned in NUhEG 0B00 SRP 6.2.4 (3).

Could you provide documentation from the NRC which indacates thear concurrance that the DTVS can be amplemented and tnat such actaen doec not require a change to the Plant Technacal Specifications?

QUESTION 10:

One of the arguements for the DTVS. that the cyctem offerc

• sagn2ficant( 3) *,pr ovements r elative t o ex2 stang ver.t am capabality comes as a surprise to observers who were not aware that plans for contaament venting during severe accadents had been prevaeusly developed.

Could you descrabe the contaanment venting procedures which ex2rted before the amplementation of the DTVE. and hc+ the DTYi offers a significant improvement to that system?

Had these prior plans ever been approved by the NhC7

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L DTV51 -- 6/12/85 Page 3 l

Supsorting backup documentation which exists would be j

heaptua to our committee, such as e

Loate Diepress UTSAR/ Tech Specs 1

F & 2D's Relevant Procedures Elec. One-Lane Diagrams System Descriptaons Thank you for your consideration of this request.

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l Notes:

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A. G. Bird, Senior Vice Prest ent--Nuclear. BECO.

Directori VarIa. formation Letter dated July 4.1987 tu

.A.

2n Dav. of Reactor Pro 5ects 2/2, NRC.

Regardang Palgram Stat 2on Safety Enhancement Program' 2)

5. A. Varga, Letter dated August 21, 1967 to R.G.

Bird

'3nitial Assessment of P11 gram Safety Enhancement Program' Co1Jankb Deputy Director, Division of Reactor 3)

5. J.

Bard

1. Sb 1986 to R.O.

'NRCRegaon1,InspectionRepor(31 N

Letter dated F.a Pro $ects, 293/66-22' 4)

R. G.

Bird Letter dated August 38.1986 to US NRC, Document bontrol Desk, ' Revised 2nformation Regarding Palgram Station Safety Enhancement Program' mm D

s.

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TOWN OF PLYMOUTH II Ltneoln Sirnt Plyrnouth. Manuchuwtis 02360 (506) 747 1620 9

september 8, 1989 fo: Plymouth selectmen cc Members of the Nuclear safety,and Health Technical support Group Thomas Murley,NRC Charlie Marshall, Pilgrim Resident Inspector From: Plymouth Nuclear Matters committee This report is a translation, summary, and critique of the 100+ page Augmented Inspection Team report of the April 12, 1989 accident in the Reactor Core Coolant (RCIC) system at Pilgrim.

We hope that these pages elicit a wider public discussion of the accident and provide access to technical information for those unable to study the larger report.

We conclude that this accident was more significant than indicated by the AIT report.

Further, that certain aspects of the AIT conclusions were incorrect, the technical analysis was faulty, and the cover letter and executive summary did not reflect the serious nature of the accident as described in the body of the report.

In our review of the available documentation describing recent problems at Pilgrim, the April 12 sccident is by far the most serious.

Indeed, the number, variety, and degree of errors and malfunctions which occurred could, under probable alternative situations, have caused a far more serious accident, endangering the health and safety of the public.

l At a minimum, we are requesting that those authorities who are responsible for protecting public safety and regulating the nuclear industry at the town, state and national level study this i

accident and strongly request that the NRC convene an Incident

. Investigation-Tsam.

This higher level inspection team will not only review the details of the accident, but also, from a broader perspective, assess the influence of the regulatory process on the cause or the course of the event.

One of our concerns has already been realized when NRC commissioner Zech responded to Alba Thompson's letter of July 13, 1989, stating, "(the event) was evaluated by the AIT to be of l

minor safety significance with minimal effect on plant equipment".

These conclusions by the NRC must be challenged, for l

l they are not supported by the facts of their own investigation.

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our additional concern is that now that the enforcement action has been issued, the matter will be shelved; the scrutiny of both specific and generic concerns will cease and necessary corrective actions will not occur.

An annotated version of our report is available for those who wish to study in greater depth the full AIT report. Should further clarification be desired, please write the committee care of the Town, or conta:t committee member David Dixon and 508-946-1000, ext. 2497, during the day.

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0 Plymouth Nuclear Matt s Committee l

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s RCIC REPORT -- 4/22/89 PAGE 1 i

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SUMMARY

AND COMMENTS ON THE APRIL 12 1989 ACCIDENT AT THE PILGRIM WUCLEAR POWER STATION i

PLYMOUTH, MASSACHUSETTS er

==

Introduction:==

On April 12 1989, during a test of the Reactor Core Isolation Cooling (RCIC) system, radioactive high pressure water backed up anto low pressure piping systems.

causing damage to equiement and the release of radioactive water and steam into the RCIC Area and the Residual Heat Recovery Area B (kHR-8).

The-accident was caused by an unanticipated combination of errors by several different people, errors in approved procedures, and by faulty valve maintenance.

It was an event which could have caused an ' Interfacing l

Systems Loss of Coolant Accident (LOCA>' a scenario where the cooling water leaks out of the reactor.

This type of LOCA is particularly dangerous because the containment is bypassed.

soston Edison reacted very well to this event and the NRC too):

keen interest, dispatching an Augmented Inspection Team (AIT) to study the accident.

The types of problems which occurred could have, under credible alternative condiatons. caused far more serious consequencen.

Yet BECO concluded, and the NRC concurred. that the accident i

was not even an ' Unusual Event.' a classification which andacates merely that the level of safety at the pJant had been decreded.

More disturbing. the AIT concluded the accident was not a significant precursor to en Interfacing Systems LOCA.

There are many disturbing aspects to the April 12. 1989 accident and the subsequent NRC riport.

The purpose of this summary te to evaluate the accident, and translate the AIT report.

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II. What is the Logic System Functional Test (LSFT) for the i

Reactor Core Isolation Cooling (RCIC) system, which was being performed when the accident occurred?

The RC2C is a safety system wh$ch provides another means of supplying cooling water to the core during certain i

accidents.

It backs-up the High Pressure Coolant Injection (HPCI) system, serving a similar function.

However, the RCIC as not taken credit for in the safety analysis of design basis accidents, so it is not considered an Engineered Safety Feature.

The purpose of the LSFT is to demonstrate that the RCIC pump shuts off af the reactor water level gets too high, but automatically restarts when the reactor water level drops to a preset i

Iow level.

The RC2C LSFT (procedure 6.M.2-2.10.11.1) is done svery six months. as per the Technical Specifications (TS).

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III.What happened before and during the accident on April 12.

19897 Prior to the accident.this Prior to the accident.

was supposed to happen:

this happened:

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.........es All involved This was not done.

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RCIC REPORT -- 8/22/80 PAGE 2 f

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on this infrequently done l

l procedure.

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I The control room operator Evidently done correctly sets the position of eight The report does not l

valves, actually changang indicate problems.

two of them f rom closed to j

open.

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r An operator was to position.

Of the seven, six were the circuit breakers for the positioned wrong, and i

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power to seven motor one valve, not even part i'

operated RCIC valves, either of the test, was turned on or off, and put tags on off because of a typo an i

the circuit breaker handles.

the procedure (1301-27).

j A second operator is to The two operators had check the work of the done the tagging first operator.

together, and acparently did f

not check each others work.

l The Instrument and The I&C technician Control technician, who signed the sheet.

I was running the test, was to review / inspect and accept the tegnino and j

sign the tegout she,et.

l Thw control room operator They did not observe the and the I&C technicaan problems indicated on should have seen from the graphics panels.

the graphics panels in the control room that I

the valves were not set up correctly for the j

test l

The logic test was then begun, involving the lead I&C i

technician, the control room operator and two other ILC l

technicians at a control panel in another part of the plant.

Durino the test, a restart of the RCIC is simulated, but with power to the RCIC pump blocked off.

But since the RCIC pump discharge valves 2301-49 and 1301-S0 still had power to their actuaters (incorrectly),

they opened.

Since the upstream side of those valves was not pressurized, water backed up into the RCIC pump and the RCIC pump low pressure suction piping.

Check valve 1301-50 1s supposed to close, prohiniting flow in the l

upstream direction, but it could not, because it had been i

matershi p(reviously) temporarily repaired with an injectedand when the valve had been subsequently Furmanite,

Inter overhauled, some of the Furmanite was left on the i

prohibiting its closure.

Not, high pressure valve,elen, backed up in the system damaging some water tnus instrumentation, opening a relief valve, and causing thermal and pressure shock to the RCIC system.

The reisef l

valve spewed radioactive water and steam into the RCIC e

area, and since the floor drains are connected, the i

Ree1 dual Heat Recovery (RHR-B) area was also contaminated i

with radioactive water and steam.

IV. What went wrong?

The personnel did not follow procedures for the RCIC LEFT

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a' RCIC REPORT -- 8/22/69 FAGE 3 and did not follow procedures for positioning the valves or for tagging circuit breakers.

p The control room operators did not recognize nor correct the problems shown on the system graphacs panels.

Ultimate responsiblity lies with these senior individuals.

who failed, in this instance, to perform their duties.

The k'nowledge available from a similar 1983 accident in the High Pressure Coolant Injection system was not incorporated into plant documentation.

Even thouch the written, approved LSFT procedure contained a critical error, somehow the error had none undetected during previous, supposedly uneventful LEFT's.

l The control of the Furmaniting procedure was poor, as war

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the subsequent check valve overhaul.

When valve 1301-17 was tagged out in the open position, i

the plant lost redundant containment i s ol a t t e r.,

in violation of the Technical Specifications.

This an itself requires notification under 10CFR 50.73 and possibly the i

declaration of an Unusual Event.

It is unclear to both BECO and NRC whether or not the check valve 1301-50 is a containment isolatson valve a r.d if so, that it should be leak tested as such.

The AIT tabled this issue to ' future FSAR revisions'.

t It may have been discovered that the leak testing elsewhere,perhaps fer 1301-50, and procedures for the check valve other check valves at P11erim and do not indicate the valves actual leakage when installed. Further i

study is pending.

V.

How did BECO respor.d?

The Augmented Inspection Team's report indicates that the BECO 1mmediate response was appropriate and timely.

Specifically. the radiological protection organization's response to the event was prompt, efficient, and thorough.

Eleven people were slightly contaminated.

l After the event, BECO formed three investigative teams,

-i led by an oversight groups a team to evaluate the effects on the RCIC system, a team to detail the accident, and a peer review.

VI. How did the NRC respond?

The NRC's William Russell. Region 1 administrator.

initiated the Augmented Inspection Team on April 13, 1969.

Their report was published May 8, 1989.

The AIT is NRC's second level events investigation the first level being an Incident Investigation Team (IkT).

It should be noted that the convening of an AIT for an event deemed by the licensee to be less significant than the lowest level Emergency Action Level - ' unusual Event' -- possibly indicates that the NRC felt that the event might have been more serious.

Perhaps the reason the NRC took great interest in the event, was the possability that this accident involv0d an

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l RCIC REPORT -- 8/22/89 PAGE 4 t

P Interfacing Systems Loss of Coolant or was a si l

toenInterfacingSystemsLosschnificant precursor event Coolant Accident (Intersystems LOCA >

1 Criteria which exist within the NRC for the determination of a sagnifacant event ares 1.

Event sequence not prevaeusly analyzed or could be far more serious vath credible alternative I

condations.

2.

System interaction resulting from a previously unrecognized interdependence of systems and l

components.

3.

Improper operation, maintenance, or design that has cr could cause common cause/ common mode failure of a safety system.

4.

Unexpected system or component performance wath sersous safety amp 12 cations or radastaen release.

5.

Multiple failures (including personnel errors) occurred an the event.

6.

Equipment failures (particularly non-safety equapment) that caused sersous transaents and j'

challenges to safety system.

A case can be made that all of these conditions were met.

The April 12, 1969 accadent at P11 gram was very signa scant.

If the AIT report is studaed closely, other probleme are uncovered which are not discussed in the cover letters, executive summaries, the Licensee Event Report, or the news summaries.

Tarst, it is not known for certaan when the event terminated, or when the the release stopped.

Second, it is not known how much water backed up past the check valve 1301-50.

Third, it is not known what pressure was seen by the RCIC pump or suction pipino. Epecifically, the logac used to arrive at a figure of 400 osa was l

Ancorrect.

The fact that the pressure switch 1360-21 wac not ruptured does not indicate that the pressure remnaned l

below 500 pai.

Rupture of the switch can occur in a rance of 900 to 2000 pai,lly occurred.y unreliable andacator of and is a ver I

what pressure actua Fourth since the duration of the release as unknown and tbe pressure of the p1nino as unknownlso unknown.

the amount of water released by the relief valve as a The 'approximately 100 gallons' referred to in the AIT report is optimistac guessing.

VII. What'1's an Interfacing system LOCA?

l In NRC's words,'Recent BWR operating experience indacater that the pressure isolation valves may not adequately protect agaanst overpressurazation of low pressure i

systems.

This overpressurization may result in the rupture of low pressure pipang.

This event, if combanec with failures in the emergency core cooling systems (ECCE) and other systems (eg. feedwater) that may be used to provadc makeup to the reactor coolant system could result an a core melt accident vath the possable release of

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7 T

RCIC REPORT -- 4/22/89 PAGE D l

fission products outside the primary containment.

Some ECCS failures may be she direct result of the an gtal rupture and/or its environmental effects.'

T This type of accident, at should be emphasized, as very critical because at bypasses the contaanment and at bypasses emergency preparedness.

It as a ' hot' topic in NRC circles (see attachments).

Two recent Interfacing Systems LOCA precursors have been scrutinized:

a January 20, 1989 accident at Arkansas Nuclear One Unit 1, and a March 8, 1989 accident at Vogtle Unit 2.

Furtherbolling Water Reactorthe NRC recently asSved,NUREG 5124' Interfacing Systems as (see LOCA:

attachment).

This report is mentioned in the A27 report, but it is unclear whether the AIT report is securate.

The AIT report and1 cates that BECO complies with the recommendations of NUREG 5124.

However, SECO's Technical Specificatione re9uire en nCzC LSrT every six months and one of NUkEG 5124 a main conclusions is to perform this test only at shutdown, when the reactor $s depressurized, an order to reduce the chances of an Interfacing Systems LOCA.

For BEco to comply with NUREG 5124, a change so the Technical Specifications would be required.

VIII. Was the April 32. 1989 accident a potential precurser to an Interfacing Systems LOCA 7 The AIT report argues that there were several isolable barriers in place to avoid an intersystems loss of coolant:

the check valve 6-58A, the check valve 1301-50, and the two block valves 1301-49 and 1301-46.

First, check valve 6-58A is a 'feedwater check valve' which as known for frequent failures.

In particulst, leakage test results for Valve 6-58A are very poor.

And were done toda based on past history, if a leakace test it is likely that it would fall. Next relying on 1301 y, s0 is questionable because it is not cerkaan that the valve ever closed during the accident.

Finally, valves 1301-48 and 1301-49 vere involved in multiple personnel and administrative errors they were incorrectly described an the LSFT procedure, they were incorrectly tagged, improperly verified, and not observed properly in the reactor control room.

To base the analysis on the adequacy of these valves, is overly optimistic.

The AIT used a variety of narrow criteria to avoid concluding that this was a Interfacing Systems LOCA.

Yet the NRC has recently said that that type of analysis is not proper and does not help achieve the geel of reducang the vulnerability of nuclear power plants to Interfacing System Loss of Coolant Accidents.

IX, Could it have been worse?

There are many credible alternative conditions which would have made this event much, much worse:

--The could have been operating at a higher power level. plant

--Check valve 6-58A might not have been able to prevent backflow.

--Check valve 1301-50 might have stuck 40 or 60 degreer C

~7

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RCIC REPORT,-- 8/22/89 PAGE 6 9

1 off its seat, rather than the assumed ab degrees.

--the opera +. ors might not have concluded that the correct action to takt was to close valves 46 and 49.

After all, these valves were supposed to have been closed.

tagged, with power removed from the motor operators, making them inoperable from the reactor control room.

--the low pressure papano could have ruptured.

--the steem release could have degraded the environment at both the RCIC and the kHk-B to the point where these systems would not be available to help maintaan adequate coolant level in the reactor core.

The AIT report did not include en evaluation of the potential consequences of credible alternative condat1one.

en important step an a well executed analyras of thas potentially disastrous event.

It is not known why.

The analysis by the AIT did not even share the concern

... the er r or s ar.d evidenced by SECO's conclusion that programmatac deficiencaec noted could have caused i

significantly greater problems under other circumstancec.

i Therefore, this event should continue to be treated as J

significant."

]

X.

Several things need to happen to resolve the issues raised l

by this accacent

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The check valver 1303-50 and 6-b8A should be leak tested.

The RCIC LSFT should be redone (procedure 6.M.2-2.10.11.1 j

The RCIC damage evaluation should be closely reviewed by independent technical experts.

The design problem concerning the placement of the check valves and clock valves should be resolved.

Analyze the NRC enforcement actions for appropraetenese.

Resolve the classification problems, and asse:aated testing requirements for the check valve 1301-bO.

Review SECO's compliance with NUREG 5124, and change the Technical Specifications as required.

Convene the hagher level NRC events investigation, the 2ncident Invest 2gataon Team ( 21T ).

The difference free I

this and the AIT is that the IIT is broader in scope, and l

includes an evaluation of the influence of the regulatory proceps-en the accident.

This serious request is made j

necessary by the type and degree of error an the AIT analysis, the contanual problems which are occurring at P1Jgram durano the ongoing power ascension program, the vader implications of the root causes of thas accadent for management of the facility, and the closer scrutiny required by a fac21 sty which as one of the worst an the nation, by several ob]ective measures.

XI. Conclusions The April 12, 1989 accident at Palgram has serious implicataons which were not thoroughly evaluated ner objectively reported in the NhC's Augmented Incpectaon Team report.

It could have been much worse.

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.i' RCIC REPORT -- 4/22/69 FAGE 7

o The chorecter end number of causes of this occaddht may be i

are deepl Further NRC investigation is warranted.y disturbing.Independant assessment of i

unprecedented and certain technical aspects is also warranted.

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Furthermore, when this accident is vnewed in light of the other' problems which are occurring during the power i

ascension, the SCRAMS, the paintenance and design oroblems, the unresolved va'1ve actuations, the equipaent le11ures, the personnel errors, etc., at seems prudent to question whether the intense pressure to get Pilgrim bach on line is contributing to en unsafe situation with potentially disastrous consequences.

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Manuseript ed: Petwvery 1W Oete Publehed:

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9mpered by T.L. Chu, s. Stevenev, and R. Paapetrick l

Conanbutere J. Lehner, Appendix E.

i A. Mngle, Appendix I union 1

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l Propered for l.

Division of Safety issue Resolution l

Offloo of Nuoleer Regulatery Research U.S. Nuolear Regulatory Commission i

Washinsten. DC SM NRC PIN A3 Bat l

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MURLEY LAUNCH 88 STUDY OP RISK OF INTERFACING SYSTEMS LOCAs Thomas Maley, duecer 'of NRC's Othee of Nucles Reseser Reguneden, has launch d a potam a I

senare that pebabilistic reek assessmenu (pR As) assumsty mAest the low pebebility het an iniarfu.

ing sysums less of seelant notidsM (LOCA) will lead is asvare core melt accidsats wins 6igni& cant off.

i she seleases.

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SWDe NA4..-assel30 som is Mwley initiesed the effort because he is shoptieel of PRA dem Wet valformly shows abs chantes of l

envede sore melu pomped by lawffacing sysums LOCAs are now. "I have se be frank,"Murley and die Advisory Commites es Rearter Safeguards (ACR8) Apnl 6. "I am not believing $e numbers. The i

aambers em taliing us thal.Jalersysum LOCA is att a problem. I shesten't any I don't h L11 eve IL I say

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l's shapdoel, se ee'se going a sert sking some andens."

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The sensept of as inimfacing symem LOCA-en essident esquenos beyond design % t zu Ars l

I idendeed la the Ip?8 Mah.1400 seaster afear giudy, and mes labeled PWR esquence V.

Mwley aid preswoor events a the intersystem LOCA econano-4msteding the 1937 q vem at ths l

%st German Biblis. A PWR--eencerned him and promped him e imidais the NRC mvier program (INRC,5 Dec. '83, I). Linder the esquenos, initure of the check valves espuedag the prise wy circuit tem the low.preeswo hQastian system perden of to emergency core eetling symem souhl result in a LOCA shs sudder.ly diosharges into the low posess eyesem and bypeseos senseinment 7hs NRC inlaissive mes only ebein "a week end" when Marley addressed shs ACRg, e nd he said he doesn't endolptu regstring any sessies indusry inidailves at this sins.

"%s goal is to save high canadenes-ed I strees that high assadeaw-ehat the poh tbility of an I

inserfesag symems LOCA, which could lead a en vmestable LOCA outnde eenminnent a less than sen te.es. mines.sia per year for each plant la the UJ " Mwley said. Mwley added that iIRC hopes a weep up the veview is about a year, and ihm tbs agency may, depending en es eussene of the review, sesammond ebenges is ladustry saining preparas a see wheter shey esa bs improved se that operators will be "sensidaad" to the signl8eanes of ths long. postulated assident esquence.

i

( how cleos they had eems or what the ramiAcadens were of the citanden, so hele induary At seestors that have emperienced preeweers a the esquenos V event "she operators..didn't knew as well as NRC has a he sensistand." Mwley esia.

i "This esquenos is imponent in my judgment beenues it bypasses abs eenenlaineat and h bypasaw emergency preparedness," Mwley said in esfonding his decialen se move derward wie the inlain'%. "It offeedvoly bypasses two levels of ew estense in.dspih enfaty philosophy ender the worst drcussatan.

est," Marley enid. "he worst cirsammenses (are) that you have a break out in she RNR (osidual heat somoval) tysiam which then causes you is not only less ecolant but m iens all yew esfoty Weeden

• capability, and which slumainly men leads a sore damage and eere solulows to en open aenalament.

"Thu goes artight le the Mmesphate and it can happen in a shers tims." he added. " Tbs wors time a

L entsuladens that I've ason can lead se ears uncoverage in a half how, ears damage in 45 m notes, and eff alu doses la the 100 rom range in an hour er hour end.6. half. So k's the imperiones of Ant esquenes shu saused me to sensider taking another look at it. I have no tv6dence that the protability of it happen.

lag is higher aben whu is emid in the PRAa, (but) I'm naardng to ens thens preeweers, es rahat than sake the PRA teaults at face value,I'm going to be a liule skepusal,just becanes of this esqueasi and lu son.

aqueness."

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Mwley rejecihd suggendens by ACRS members shat the esquenos V esonene b' eensie ired u pan af es Individual Plant Eaaminadens (! pes) $at NRC has seguired of all UJ. nuclear facil nes.

"I eink k's just going to overbwden 175," he es!d. "lPI was never meant to to ths vehicle is seselve all lesues asseeisted with asvere accidents. If we were e ask licensens to leek at av mt Y as pan etabstr IPEs, three years from new we would get back esmeming that I almost gusantes = ouldn't be earth anything. I don't think they have the methodelegy (Ast) would be good enough (ao) Aas I would be stiansd and I sleo don't want to wait for eras to ove years."

Last year, when details of the 1987 Siblis svent swfusd. Mwley said the agency was considenr.;;

the need for further guidanse on the issus.- Osw han. WasAnsten Sn Mi%,M-SM pul a n M er

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This study wee perfereed by the Risk svaluatirn Group, tapartiqtat of Waele-er Energy. Broekhaven Motional Emboratory for the Office of Nuclear Regulatory Research, tsaster end Flaat Safety !ssues Brasch Mvielen of taaetpr and Plant i

Systems, U.S. Butlear Regulatory Cossission. he objoettves of that study are to tievoettgate the vulnerability of eartent belling water reaster (WR) designs l

l go at taterfestsig systems LOC & (ISL). identify any improvements that would sig.

l tittaastly reduse the frequency of 181a. deterstae the sost-benefit considera-i I tiens thereof, sad detersiae the effects and the seet benefit relationship of tastituting teak testing programs of the pressure toelation valves for these (plante that de met surrently have such a requirensat.

i ate stady 1. based.,oa the d.taned e.a.inaues of three,1.xs c,eaa tottee. Else liste Point 3, and Quts Cities) with the goal of taking the plant-opecific findings and entrapelattag the results to aid la the rese14 tion of h'RC

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= o,.r. stag o,. u.c o sadie.t.. aat no,,e...e i.,1stie.

l valves may set adequately protest against everpressurisation of low pressure l

erstems. This everpressurisation any result in the rupture af low pressure

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i-y pipias. This event. if enobined with f attures in the eastgency cort cooling

.4 erstems (ECCS) and other systems (e.g. feeduster) that any be used to provide

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.a.u, t. the v et., co.l.at.y.t.a. seu14 r.sutt in a e.r. melt a teid.nt.ith l

the possible release of fission producto outside the primary contai nment.

Some j

l SCC $ fattures may be a direct result of the initial rupture and/or Lts i

j environmental effects.

I One of the primary seals of this study was to determine the cost-benefit

(

telattenship assestated with requiring plaats that de est surrently have leak testing requireeests on their pressure isolation valves (PIVs) to t natitute such a program, Bewever, all of the reference plaats already have vario4s require-meats related to leak testias.

Therefore it uso decided that since none of the reference plaats represented a true " base case" model in this area in additional base case model would have to be created.

The base ease model was taken to be the peach Setten sedel with the p1T 1eak testing aspects removed.

temoving the leak testing benefits free the peach lettes model resulted in a large increase jee, predicted eere damage fre(senty due to Ist.

in Essed upon the resu Lts of a

-ate sensitivity study. it a,,..re e.ffineo for as lea tasung,regr..

1 to include prettetens such that leak testing be performed at each rsfueling as l

lwellasaftertedividualvalvemaintenance.

he risk-based benefiti calculated for thie leek testing progran show that euch testing schemes are soit i

affeettve.

la addition, the of fsite risk-based cost-benefit considerations for the suggested testing program were calculated to be fully test effectiv.6 whether er not the break in the low pressure systes was assumed to be subserse1 under water. A submerged break would result in trappina of sese of the a presol fis-l sies products in the water and thue lower the predicted of fsite con sequenceb.

The results indicate that in spite of uncertainty in predicting fisaien product release the benefits in risk reduetten outweigh the cost of impleoesting such a J

1eak testing prograu.

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he tasights free this study fall into two basic categories.

The first aategory deals with essuring that the pressure boundaries are intact prior to asseenstag reester pressure and the second category deals with key te aveld l

plastam the plant,eanecessarily into a more vulnerable mode of plant operation.

i table

, provides 's convenisat sellection of, the. pertinent tore desagt freguan-esos (CDye) presented throughest this report.

Se table will be user to facilitate eenparisons and derive 17. sights.

Se first category obeve is addressed by p1V leak testing provie tons.

From Table 1, " peach lettes (ao leak testing)" represents an analysis whos ein the Peach Settee model was stripped of all credit for its current leak testing prac-i l

" peach Bettee (current)" refers to the peash pottes plant es found and tiees.

andelled.

"peseh Bottom (with leek seating)" reflects the niataus leak testing previsions derived free this study (i.e. leak testing all air-operate d check valves et each refueltag and ladividually af ter malatenance).

Compasing the "no-testing esse" to "pesch lettes (currest)" shows that the esisting level of leak testing has already reduced the peach lettee CDT due to IgLs by an order of j

eagnitude.

Comportag " peach Botton (current)" to "pesah Settee (with leek i

t: sting)" shove asether order of magnitude reduction is still available.

A f

significant benefit (similar to that derived for peach lettes) for such a leak testing program is espected to hold across the SWR population.

j the second category of tasights is addressed by changing current testing practices.

These testing practices can be almost as significant as 1splements-tien of a leak testing progras, however, they are tutte plant-specific.

The I

4:sinant osemple from this study is fesad at Nine Mile point 2 (NNp).

By I

temparing the two imp-2 entries in Table 1, there is apparently more than a two

[ order of magnitude decrease in the CDF for 1st available by prohibiting the t

currently allowed practice of stroke testing the valves in the steen tendensing i

lines to the ENR haat eschengers (with the reactor pressurised) and allowing the stroke testiad to avait a eenvenient shutdown (with the reactor depressurised).

i A second example of significant testing-induced risk can be seen by compar-ing " peach Sotton (curreat)" with " peach Sotton (logic test at shutde m)" f rom Table 1.

This is the single most effective corrective action identif Led for the peach Bettes plant'in reducing sore damage frequency.

Current peach bottom testing requirements include the provision to test the 20C3 logie eva ry six esoths ladopendent'of~whether or not the reactor is pressurised.

By solding off en the ECCs logic system functional test until a reactor shutdown cosas along, (i.e., the reactor is depressurised), the 1st C0F can be reduced by a ;sost an Crder of angnitude.

In summary, the results of this study show that institution of a etnisus leak teeting program for the air-operated pressure isolation check va,ves represents a significant reduction in the estimated ISL CDT for the three plants studied, which should apply across the entire BWR population.

In addntion it has been shown that some of the current SWR teettag practices can also represent o large contribution to !$L CDF and that this testingainduced risk is easily removed by rather simple and cost-ef fective changes to existing testiog procedures (as discussed directly above).

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Plant State CDF/ Year 1

Peash letten (We leak testin4) 1.865-5 Peach lettes (Current) 1.028-6

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Peach lett.n (tfith leak testing) 1.975-7 O

Wine Mile Point I (Current) 4.41t-6 i

Nine Mile Point 2 (With all fitte) 3.225-8 I

Peach lettes (Logic test at shut 4evn) 1 215-7 i

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