ML18038A719

Forwards Petition for Emergency Enforcement Action & Request for Public Hearing Relating to Facility Operation in Violation of NRC & Federal Requirements for Availability of ECCS High Pressure Core Injection

ML18038A719
Person / Time
Site:	Nine Mile Point
Issue date:	10/27/1992
From:	Ridings B AFFILIATION NOT ASSIGNED
To:	NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
Shared Package
ML18038A720	List: ML18038A719 ML18038A721
References
NUDOCS 9211160396
Download: ML18038A719 (76)

Text

MEMO FOR YOUR FILES Oc t 27, 1992 TO:

U.S. Nuclear Regulatory Commisson Executive Director for Operation-Public Document Poom l7i7 H Street Mashington, DC 20555 FROM:

Ben L. Ridings P.O.

Box 110i

Kingston, TN 37763 Pef:

Petition pursuant iOCFR2.206

Dear Sirs:

Enc.'osed for fil ng PETITION OR EifERGENCY ENFORCEMENT ACTION AND REQUEST FOR PUBLIC HEARINC.

Pespectfully submitted, Ben L, Ridings C

g

UNITED STATED QF AMERICA BEFORE THE NUCLEAP. REGULATORY COMMISSION PETITION FOR EMERGENCY ENFORCEMENT ACTION AND REQUEST FOR PUBLIC HEARING 4

I.

INTRODUCTION I,

BEN L.

RIDI!lGS (hereinafter "Petitioner" )

hereby petition the Commissioners of the Nuclear Pequlatory Commission

("NPC" or "Commission" )

for emergency enforcemer<t action again t Niaqra Mohawk's Nine Mile (Unit One)

Nuclear power pl ant, which i s operating in vi e<1 ation o

< both the NRC and Federal requirements for avai.'ability of Emergency Core Cool.ing (ECCS) high pressure core inJec t ion.

As an ECCS system, the I! ne Mi 1 e pl ant also fai 1 s to provide the

<ra<ndatory emergency backup power to ':he high pressure core injection (HPCI) system Over the twenty years the Nine Mile One plant has been allowed to

operate, no safety related pumps have ever been available to inject water into

,he vessel at reactor pressure.

A. t"e same time this plant was allowed to

-perate at ful? powers here a(e mal<p postulated accidents assumed in the Final Sa<ety Analys s Report (that ar capabic of draining the reactor vessel) and specific.~'.?y rely on the ECCS HPCI Pumps to maintain reactor water level.

These pumps have never been installed and the urrent administrative controls al?owed this plant to operate outside the m'lnimum federal requirement.

This specifi=

type of plant operation outside the kr<own;<'.<niwum federal requ rements greatly endangers health and property risk to ".he:.'@bi ic.

As discussed in detail

beiow, the responsible utility, its Quality Assurar<ce group and the NRC have routinely failed in their responsibility to ensure the cperatior. of nuc'ear when problems arc identi fied, power p?an".s within the license agreement.

Even documented are rcuqht to the at'tention of responsib.e

parties, various safety

oncerrs

re routinely dismissed, ignored or

~

~

"administratively eliminated.

Even issues which obviously endanger public safety have been routinely dismissed, not only by the utility but such actions authori"-ed and approved by the independent quality assurance groups and by the NRC.

Any and all of these organi"..ations have the authority to stop the operation of plants outside the minimum safety requirements, and not one have come forward to fulfill i ts duty and pr otect the public.

Instead, each organi "ation has reviewed the enclosed safety concerns and contrary to any practi =al Justifil ation, have remained si'ent and allowed this manner of plant operation to take place with th~;:

" -"~q',

aivinq evidence that these groups have also failed to remain independent of each ether.

Independent review by not onl j~

the government agency but the quality assurance review groups is the basic premise which allowed congress

".o grant operaticn of commercial nuclear power plants w.'th limited liability d'or damages, The =urrent administrative controls u"ed today failed to ensure the plant operate within the minimum federal qu'delines.

It is Congress's duty to protect public safety and its current administrative controls have fai'. ~d.

because the Nine Nile Point L'nit One Reactor v olates both federal law and the Commissions's requirements fcr O'GH PRESSURE COPE INJECTION the Commission can make no finding that,~here is resonaole assurance of no undue risk to public health and safety.

Petitioner therefore request that the Commission issue immediately an effective order direct;"ng the licensee to cease power operation and place the reactor in a cold shutdow: condit;on.

The plan"; should not be perm tted to continue or resume operation unless and until subsequent tests and inspections are shown to provide the requ's:te reasonab.'e assurance of no undue r sk :o p blic health and safety.

i oreover, P titioners seek a public hear!ng before the p!ant i-allowed to operate

"-;.in.

II.

DESCRIPTION OF PETITIONER I,

Ben L.

Ridings, am a technical consultant for commercial nuclear power plants.

Over a span of some fifteen years, while ~orking at some twenty four

sites, I have speciali=ed in reviewinq of licensing agreement (FSAR, Technical Specifications, Federal Codes and Requlations, ASNE Codes, etc.), establishing administr ative controls to meet these requirements and test programs to ensure compliance at all times.

Ny test programs and administrat i -

controls established while under contract to various utilities are still in use today at many facilities.

III. THE CONNISSION SHOULD EXERCISE ITS SUPERVISORY JURISDICTION OVER THIS PETITION A.

The Commission has an Inherent Supervisory Jurisdiction over the Safety of Operation of the Niagra i'~ohawk Nine Nile Plant.

This petition is brought before the Commission pursuant to the authority granted to it in 42USC

'233(d),

2236(a),'37 and 10CFR 2.204,

=.206(c)(i),

50. 54,

50. 57,

50. 100 and 50. 109.

It invokes tt>e inherent supervisory authcrity c f the Commission to oversee a'll aspects of the regulat'ory and licensing process and

-t=

"overriding responsibility for assuring public health and sa!.=ty in the operation of nuclear power facilities."

Consolidated Edison Coo.

of N.Y.

Inc.

(India:, Point Units 1 p 2 and 3)

~

CLI ?5 8y 2 NRC 173

( 13?5)

~

As the Commi ssi on has previ ousl y

observed, its supervisory pcwer s inc.'ude the power to order immediate shutdown of a

facility "if the publ c health or safety so requires.'etition for Emer enc and Remedial Action, CLI 78 6, 7 NRC 400'05 (1378 g

citinq 5 USC 558(c),

42 USC 2236(b),

10CFR 2.20 '(f), 2.204.

lf

The Commission has exercised its inherent authority on a number of occasions.

In addition to the ceases cited

above, see Petition for Research and Devel o ment Administrati =n (Clinch River Breeder Pe."-.c". "r Pro'ect),

CLI-76-13, 4

NPC 67, 75-76',1976);

Consumers Power Co.

(Midland Units 1 and 2),

CLI-73-38, 6 AEC 1084 (1973);

Public Service Co.

of New N~ae shire (Seabroot Nuclear Poser Station, Un t-i and 2), CLI-77-5, 5

NRC 503 515-z'7'1977)

B.

Exercise of the Commissions's Independent Jurisdiction s Appropriate n Thi-Case.

NPC regulations at 10CF'R2.

06 provide that under ordinary circumstances, enforcement petitions are to be lodged with the NPC

Staff, and that the Commission may take discretionary review of Staf f denials of such petitions.

However, the Commissions's reviewinq power "does not limit in any way" its "supervisory power over delegated Staff actions",

10CF'R2.206(c)(1).

It is appropriate for the commission to exercise its supervisory powers and take jurisdiction in this case because the NPC Staff has acquiesced to Niaqra Mohawks'iol ations for more than twc years.

In Jan 19907 Niagra Mlohawk Compliance Supervisor was given written notice of HPCI and other inadequacies which ef feet public safety.

After no apparent

action, the Nine Mile Quality First Team was also given notice.

Petitioner was later notified by the Qual;ty,=irst Team that the NRC had been contacted and made aware of the problem.as well.

Petitioner was later contacted by the Quality F'irst Team and told that the NRC had exempted the plant from the HPCI requirement and its need for backup power in the event of loss of power.

Petitioner has yet ';o hear directly from the NPC on th.'s ma. er.

4 4.

1

IV.

GROUNDS FOR ENFORCEMENT ACTION A.

Federal Pequirements for having radioacti;=.'ls on site

.In accordance with 10CFP50.10, the utility Niagra Mohawk entered into contractual agreement with the federal government under the provisions of pubLic document 50-220, on file with the federal register.

Now under the jur sdiction of 10CFR50, App.

A (General Design Criteria),

establish the minimum requirements for the principal design for water cooled nuclear power plant.

Criterion 33 and 35 (Attachment 2) specify the minimum need that a system to provide abundant emergency core cooling shall be provided.

The system safety function shall be to transfer heat from the reactor core and must have suitable redundancy in components and on site electric power system (assuming offsite power is not available) which will enable the safety function to be accomplished.

Also (Criterion 33),

a system to supply reactor coolant makeup for prot ction against small breaks in the reactor coolant pressure boundary shall be provided.

Criterion 37 provides the testing requirements of the emergency core cooling system.

10CFR?0 details the utility and NPC responsibility for testing and inspection of these systems and

'10CFR50 App.

3 (Qua.'ity Assurance Criteria) details the Quail ty Assurance Progr tfll clAd the admi nistrati ve requirements for Inspections, Test Control, Operating Status)

Corrective Action and Records.

,8.

A Study of Contractual Agreement (docket 50-220)

In accordance with 10CFR50.34, the technical speci fication shall perform an evaluation of the safety effectiveness of providing for separation of high pressure coolant injection (HPCI) and reactor ccrc solation cooling (PCIC).

This 'nvest qation found the Nile Mile Point Technical Speci fication in =ompl";ance "th this requirement.

Technical Speci fication 4.1

~ 8(Attachment 3) g'.v. - positive proof that the ECCS

0

requirement for the HPCI system was anticipated by the designers.

Secondly, the corresponding Limiting Condition for Operation (LCO) 3.1.8.c (Attachment

3) view this system as so -ritical that if "the utility fails to verify HPCI operability it wi!I demand an orderly shutdown be initiated t

within one hour.

When on.'y one HPCI component becomes inoperable its redundant component shall be demonstrated to be operable immediately and daily thereafter (as opposed to monthly demonstration)."

In accordance with the Base" for Technical Specification 3.1.8, the HPCI system is provided to ensure adequate core cocling in the un 1 ikel y event of a reactor coolant:ine break (also a

federal requirement-design criterion

33)..The HPCI system i

required for line breaks which exceed the capabili y

>f the Contra.'od Drive pumps and which are not large enough tc allow fast enough depressuri;ation for core spray to be effective (cor e spray

"'350 psi as opposed to

.'-.'PCi 2200 psi ) ~

accordance with the

~ inal afety Analysis Peport (FSAR),

Chapter

',Attachment

,4),

the i;csign 3ases for HPCI is discussed.

Although severa!

.=v']sion have been lmpiem:.anted i;y -;he utility in order to fabricate

.-;e existence of a E CS system tv

.aI *sfy the HPCI federal requirement, its primary safety function is 'is{ed; (i)provide adequate cooling of the r =actor core under abnormai

-nd accident =anditions, (2)remove the heat from radioactive decay and residual heat from the reactor core at such a

rate that fuel clad melting wou!d b=. p;event'ed, (3)provide for continuity of:ore cooling over the complete r:ange., pmtuiated break si:es in tl e pr imary system process barrier.

"-n;.e ".he safety functions are understoc'd it::~"-."omes obvious as to why thi= =.y-te:...'.,

a m nimum requirement of the 4ed~i;(

. gui del ines a

The following paragraph of FSAR Chapter VII gives the reader an indication of the lack of proper review that exists.

At Nine Nile

Point, unlike every other nuclear facility, "HPCI is nob an enqineered safeguards

<system and i s not consi dered in ariy Loss of Cool ant Accident Anal ysi s."

As stated in the FSAR (in '.aye<an terms) bnis feedwater system does not pretend to meet the

!OCFR<5C Apperdix A

(Criterion 33, 35, 36, 37) requirements of the minimum federal requirements.

In fact, Nine thi le Point has no system meet ing these mll'<m

< ederal requirements.

Next, reviewing the Design oval,,ation portion of FSAR Chapter

')ll (Attachment 4) a pa< adox c<ccurs i<

design pEiilosophy.

"During a lass-of-coolant accident within the <.'rywel!,

high drywell pressure due bo a line break will cause a reactor s=ram.

<E;e automatic scram w.'ll cause a turbine trip after a five-second delay.

In order to revenb claddin tern erature from exceedin their maxi<gum limit '.or the entire spectrum of breaks the 38M gpm (from one brain of the HPC: pumps) '<~ould have to be avai!!able Qbviously,

~he HPCI -yst m ':

ab olutely <",ecessary to ensure critical heat flUx,'F) is nob exceeded

~

tv". tElou tile cool a>st water to transfer tEie

<-;eat fr 'm the fue.'o the cool ';,'.,

the fuel rod would then heat up rapidly and fue'1addinq would take place and

-ause a possible mel t down unless the reactor were shutdown quick'y.

Further, once the critical heat f.'ux was exceeded, the departure from nuc'cate b<;i ling ratio (DNBR) would exceed its 1.25 limit.

These I mits a;e

.=<hn<ical Spe" ification requirements

".s well but it gives an indication of th - interdependence of the ECCS systems.

".= make

=; "-'atemen". in a !icense

'.E<.:.:

"'HPCI has not been considered in an,

'ss of Coo.'anb Ac=ident Analyses" is

~

'.n ~ther indica".on of the lac.':

~" vi ew that exists at Nine "j'.le Point.

=very =afety limi".assumed

at. the Nine Mile Point plant is jeopardi "ed without the assurance that the fuel will remain covered at all times.

The NPC has approved the non-safety related feedwater system

'as an appropriate substitute for an ECCS HPCI federal requirement

~

What at first seems like a quibble about a

single pump is in actuality a valid argument that every bases assumed by this license is null and void.

At Nine Mile Point, standard basic thermal reactor design ha been significantly altered in several ECCS systems.

There are na HPCI or RCIC system ta transfer heat from the reactor care.

There is no way of taking steam away from the reactor and using this energy to drive a

high pressure pump.

Normally the HPCI pumps return the condensed steam (water) back into."he vessel to maintain water level.

At Nine Mile Point, there is no HPCI

r RC.'C systems.

At Nine Mile Point unlike normal eactar

design, e'ectri=ally driven, non-quality related feedwater pumps a" c co lsi del ed, These nan-qual it y rel at ed feedwater pumps supposedly fulfil'I the

.HPCI safety function and yet da not meet the electri.-al backup requirements.

It nu

". be noted that the si=e of these electrical pumps make it

<mpasslble to have on-site power available in the event of loss of aff-site pawer.

On-site power availability is assumed

'n the bases af the FSAR.

It is herefore impossible for this plant to fu.'ill the minimum safety ab.gation as dictated by federal statute of the known pc<stulated accidents.

This same feedwater system (be"ng n:n-quality related) was purchased as a non-quality related system.

In

< e) i '~

5<<<<me system;

piping, valve-,

instrumentation, wiring, electrica.'ampanents and control systems were a'.'urchased and installed under nan-qual!".<~

=.-'cntractual provision..

HPCI au.<omatically initiates on a ass Coolant Accident (LOCA) signal

'.r:~

the NSSS log.'c.

The

.VASSS log"c performs the ECCS safequard functions an~

\\

- 8

al ways installed under str i c t contr ac tual mand ates, which include training, quality assurance reviews, certified skilled craftsmen, etc.

Secondly, the piping system,

welding, hanger restraints and maintenance considerations were installed and maintained under non-quality related provisions as well.

Again, ECCS safeguard systems ar purchased, constructed and maintained under much stricter guidelines.

The feedwater system was never

designed, purchased g

bui 1 t, maintained nor capable of fulfilling the HPCI r equi r ements c f the federal gui d =1 ines.

At Nine Mi 1 e Point the HPCI system simply does not ex ist.

The admin strative controls which allowed acceptance of such a non-quality r e';ated system to fulfillthis manda.ory ECCS federal requirement is nut acceptable.

C.

Knowledge of Exi tin Concerns The need for an operab.'

ECCS HPCI System is mandatory as evidenced from the grounds for rel;of in this report.

At Nine Mile

Point, the Utility, Luality Assurance personrel and the NPC were wel) aware of this requirement.

F'r what ever reason, this plant was licensed by the NRC and al 1 owed to operate wi.'hout thi s man 'ato(y requi rement instal 1 ed.

Attempts by these same parties to substitute non-quality related feedwater equipment to ful fi:1 this mandatory safeguar d;<<nction supports the fact the need for requlremeflt was understood.

Ev'n

', non-qual ity related equipment was acceptable to support ECCS funct.'o; s (and i ts not),

there is no onsite el ec tric power system that wi 1'I suppor t the sa fety func t i on o f feedwater/HPCI system.

Thisi system is another mandatory minimum requirement (Attachment 2-Criterion 35).

.o prove the ccllaboration between all parties ment on >>d,

he 'i"ensee attempts to take credit fcr ansi te power aval I abl, I 1 ty from the : nton

Dam, some 100 miles away.

Qbvi ousl y the revi ewer s are aware of V..o e mandatory requir ements but the". o

't

0 resolution to the safety concerns is not acceptable.

The possibility of a tornado destroying the switchyard is a known postulated accident that can occur.

Without this power availability, the.,PCI function cannot possibly be assumed, as stated in the FSAP. Chapter VII (Attachment 4).

Every'ime the feedwater procedures were revised th.'s issue would have to be reviewed.

Kverytime the FSAR (Chapter VII) was

revised, the Technical Specifications r vised

=r containment integrity was questioned this issue had to be reviewed in accordance with administrative requirements set out by the ',ederal guidelines.

Everytime the Quality Assurance groups and NPC per formed their independent audits and inspections this issue had to be reviewed.

Everytime this plant was operated at modes i or the respol sible Senior Peactor Operator (SRO),

who is speci fically trained (i0CFR50 App E) vn these;ssues would !'.ave to question the validity of the current HP.I system.

Every time the HPCI surveillance (monthly) was per formed to ensure operability, the responsible SPO would have to question the validity of a non quality related feedwater system fulfilli'nq the HPC.'ystem.

Tak ng credit for non-quality related equipment to fulfill the requirements of a ECCH safety

'un. ti:n is not ac=eptable and it would be the SRO's responsibility to quest on th

'.eedwater sbili ty to perform this HPCI safety function.

Of course, tha'; is the another problem to consider, it wou.'d be the SRO'-

'~'

Although previously awar>> of th

~ pr=blem, on Jan i8, i390, the Utility was served notice of these and o'.her safety concern.

If the non-quality related feedwater system was tc supposedly fulfilI the HPCI safety

function,

.'t failed to met the onsit elec;rical requirements and many of

4

E the main flow path valves had never been included in the Inservice Test Program (IOCFR50.55).

Some 44 out of 47 valves vere currently not identi fied in the Inser vie Te.- t Program(ECCS Surveillance violation).

With such knowledge, the U i '!ity, Quality Assurance group and the NRC a.'loved the plant to start up and continue into full operating (mode 1) condition.

No

pumps, no val:es yet echni cal Speci ficat ion

4. 1. 8 (Attachment

3) demands if one, va'.ve is not demonstrated operable a

daily surveillance is required

".- be performed.

This is just another lack of administrative contrcl i r. which the reviev groups have failed to audit or

~ eviev properly.

Ur.'.'ortunately, this dilemma i s not unique to Nine Nile Point.

Other plants vere also somehow !i-ens d with-ut this 'mandatory HPCI capability.

That i:

another

'ndicator of the type of review that has taken place at other facilities as we!1 but evertually these plants installed the mandatory system.

The most st u.n ng a

o f till5 invest igati on sflows that after literally thousands of techn!cal;evievs per formed by hundreds of "quali f ed -personnel"

>> "rkinQ ir; d'.'erent shifts, separate departments, sites ol i eglonsy have all fal ed s'tc'p this facility from operating out 01 oe I

min;mum federal gu'de;.':.es.

="very month during ful 1 power oper at o.'),

the HPCI system i-veri fied operablo by a "qualified" Senior Reactor Operator ard

.", sworn a".:davit

-.',bmitted each month by the Ut. sty to tl)e NRC attestinQ that a.'.

requirements have been ful fi1led.

"b.;= ~ly, the c 'rrent system of checks ard ba'ance" cannot stop this plant

'I opera t

')Q ou't si de t )ese mandate r y

.'eder al gui de l esp an assumpt i on fal sei y

))lade by c ongi ess e

D.

Responsibilities 10CFP50 App.

B details the administrative requirements for Test

Control, Inspections, Operating Status, Corrective Action, Pecords and

ndependent Audits.

These requirements are addressed in both the Technical Speci ficaticns and FSAP..

Site spec.'fic administrative procedures detail uti lity and quality assurance staf f posi tion responsibili ti es.

10CFP50.";0 detai!

the NPC inspections while 10CFR50.72 detail report noti fication

( esponsi b i I.'..'s I or al I par t ies.: he NRC have the.'

own admi ni st rat i ve pro edures wh;ch detai! -taff respcnsibi! ities.

NUREG-0800 details the USNRC standard review plan for inserv:ce testinq of pumps and valves.

All part:es mentione.'ere required to have knowledqe of the HPCI requirements at the.'evel of review for whi=h each individual was involved.

These reviews require mandator>

act'on.

Gesp.'te all mentioned reviews this requirement was not met.

n 'an 18, 1990 th Niapra Nohawl<,

Nine

'file Point Nuc!ear Regulatory

.: mpli ance Grou~ were served not ce of this and many 2ual othe>

k"own sa ety ccncerns.

On Jul y 31, 1990 the Niagra i'lohawk

.y First Team were served written notice.

The NRC was ncti fied an" on and th ua.'ity Firs.

".earn not.'.fied pett oner that

".he NRC exempted the ut!.'ity from the requirement.

V.

STATENENT OF THE LAW There is a minimum requireme".';..'.

H "h r.-essure Core Injection CS Safeguard System at the N nl

.'-e.-o'.nt Un.'t One facility.

This requirement comes from t;.~,..', a.'uidel

nes, Technical Spec ficati =ns and r ~AR minimum man<<ates.

2.

No Hjgf Pressure C "r ~

nJectio

. -vst=m meet jnp the sa eguard fede"

"-,'u de!

nes exists at Nine Ni!e Pn"., Unit One.

r

~

~

3.

If the non-quality related feedwater system was to supposedly ful filL the HPCI safety function, it failed to met the onsite electrical requirement" and many of the main flow path valves had never been included in the Inservice Test Program.

If the HPCI System is not a safequard system and is not considered in any Loss of Coolant Accident Analyses a

stated in the FSAP.

Chapter

VII, then no assumption can be made that the fuel will rema n covered by the moderator and related safety limits set in the current license are nu'I'nd vcid.

O'iously unreviewed safety questions exist.

Congress made an assumption of the current checks and balances that would never al.'ow a p!!ant t: operate outside the minimum safety requirements set out n ',ederal guidelines.

On this assumption unlike any other industry,

'.he nuclear industry has been allowed to cperate under limited liabi'! ty.:he uti'. ty, Quality Assurance

Groups, NPC and Chief Execut.'ve Of;.'cer have received written notic

~

of their failure to camp! y wi th t"he minimum federal guidelines and have administratively failed ".- c.mply with this issue.

As discussed

above, the Nine lii'.e L'nit One Plant fails to comply with both the minimum federal and NPC'requirements for HPCI ECCS System.

Th s has been acknowledped by the

!RC Staff an" is demonstrated uncqulvocall" by the evidence in the public reco. ~.

moreover, the Staf; !las per formed l'o valid analysis that meets the Vmmlss n~ - narrow r teria for continuin-to operate in the absence of comp!iance.

power pl ant.

In fact p as the Ni~C 5 Appea!

and

'!PC safety regu.:ations

.'s u p: ereq '

te

...=."..p! iance with both Feder;~',

to safe operation of a nuclear Boa. d has ob"erved, regula"..:i~

r1 4

and safety."

Maine Yankee Atomic Power Com an ALAB-161, 6

AEC

1003, 1009(19?3).

Compliance may not be avoided by arguing that, although an applicable regulation is not me, the public health and safety w 11 still be protected.

For, once a requlation is

adopted, the standards it embodies represent the Commission' definition of what is required to pr tect the public health and safety.

Vermont Yankee Nuclear Pow Cor ALAB-138, 6 AEC 5 0, 528(1973)(emphasis added).

he Commission's essential safety standards must be met, without regard to the cost cr.'nconven nce of achieving compliance.

10CFR50. 109 See also Union of Concerned Sc Bntists v NPC 824 F.2d 108(DC Cir 1987).

VI.

REQUEST FOR RELIEF For ie (Bascns enumerated above

~ pet I0"Br st a es that the fo'I 1 owing rel i e f is required:

A.

Immediate Shutdown Pending Demonstration of Regulatory Compliance.

As discussed ab=,

the Nine !".ile Point nuclear plant fails to comply wit!1 an array of fun'>>';.men 'al

'quirements for HPCI ECCS mandatory

"~quir.ments.

No exemptions to this requirement can possibly be justi fied w thout undue risks to public safety.

Consistent wit!1 '.,',"- ".o.quirements of the Atomic Energy Act, Federa'andatory requirements and NRC regulations, Petitioner therefore seeks immediat>> sh".",d'n of the Nine Nile Point'n t one reactor pending full compliance with the regulations.

In seeking this relief, Petitioner notes that maintaininq ECCS syste.is necessary to metigate los-of c olant a": dents is a regulatory goal warrants the most immediate and str ngent nforcement action.

Nine Y'iI B Poi nt '

noncomp~ Lance

'.he f d~r=-'.'n mum desiqn criteria and cove\\'p ac,vivl tkes Gf al respan;=.!b'.B "art.'Bs which poses a safety

~ I

a W of commensurate, i f not graver, dimension than the suspicion of ECCS pipe cracking that caused the ccmmission to order 23 plant shutdowns in 1975.

See Petition for Emer enc 'nd remedial Act!one CLI 78 6) 7 NRC 400 r

405(1978)

~

Like the ECCS pipe cracking, this plant doesn't even have the

pipes, valves or pumps necessary to metigate a known postulated accident that effects known safety limi".s of the rSAR.

Th s system is necessary for the cooling of the core during an accident and this system (which does not exist) is the only means to present"a meltdown.

Again, unlike normal ECCS systems which have redundant components and ran therefore withstand a

single failure, this system dces not exi='nd cannot be compensated

or by any other system.

Simply put, a small break described in the FSAR bases as a postulated accident will in all.kelihood meltdown the reactor for lack of cooling.

Because the containment is not designed to withstand a

meltdown, such an event would prcbably 'ead to an uncontained release of radioactivity to the pub!ic environment

~

,his utility is not insured fcr such an acc dent.

B.

Public Hearing

'sues al sed by thetine Ni '

Point '

noncompliance with federal requ rements raises grave safety question" of tremendous public importance.

Petit oner therefore request that before al.'owing the Nine Nile Point ='lant to continue operating, the Commission provide for public

hearing, wi th rights of discovery and cross cxaml

~ a ion, o determine whether Nine i'li 2 Po nt is in full compliance with all;e"eral minimum requirements revelant to HPC:.

and p blic safety.

Secondly, congress be notd

~at the admlni-tratlve controls reli d

upcn to grant the nuc'lear industry the immunity ". l.'ability have ailed ensure public safety.

After literally.housands of reviews by quail tie

ii

personnel" from di fferent disciplines, departments, sites and regions completed their review, not one came forward and demand this plant operate within the law as laid out'y act of congress

~

Should noncompliance be

found, many of these revi ws demand mandatory action on the part of the reviewer.

The petitioner has ncti fied all responsible parties and after twc years Nine Mile Point Unit One continues to operate outside the federal guidelines at a

tremendous ri sk

'.o public safety.

A congressional investigation :f this matter be initiated immediately.

The petitioner' services were contracte 'y Niagra Mohawk to review and ensure administrative complian=e to Technical Specification prior to Start-Up.

A qualif:ed group of ten began a laborious review and when enormous problems began to immerge.

This group was disbanded immediately.

?n Jan

1990, the N agra Mohawk's Nuc.'ear Regu)atory Compliance Staff was given a

detailed memo (Attachment

5) giving evidence that 451 of the ccntainment isolation valves

)lad ai!ministrative deficienc'es.

Two weeks later the review group was d.'sbanded prior to completion of their review.

Along w h HPCI concerns, conta:rment isvlat.'on valves as found in the F SAP.

able V'-3 had deficiencies with corre"pondin Technical Specification Tables

-.3.4 5 3.2.7.

Th's p'lant had per ated for twenty years and yet the

'icense fa.'led to even correspord to itsel F, let alone actual p.'ant cond t'.ons.

These valves are re;;uired "y ',ederal guidelines to protect the public yet almost half had defic'cies.

Petit oner alleges that when concerns are identified, t"he =oncer;,',~re routinely "covered up" dismissed or administratively exempted.

A prcper review of the Nine Mile Point Unit One

,echnical Specification 4'... requ;rements and the comliance of the

0 test programs will show that the utility simply hired another review group that (for whatever reason) fa'.led to docunent the deficiencies that truly exist.

Nine Mile Point Unit One resumed full power operations even after the safety concerns were identi fied and documented.

This type of cover up is not unique to this plant and a congressional inv stigation of this matter be initiated immedi ately.

lN

SUMMARY

There can be no just fication for the operat',on of nuclear power plants o tside

'he minimum requirements speci fied by act of congress.

These are the minimum requirements deemed necessary b> act of conqress to qrant the immunity of liability current'y assumed by the uti ity.

Mhen public safety is jeopardi "ed by known postulated accidents, there can be no justi f ication for the lack of acticn by the responsible parties in this instance.

Simp y pu this utility is not:nsured to operate n this manner.

Respec t '.u.'-.'

submi tt ed, Ben L. Ridings P.G.

Box l.'Oi

Kingston, TN 3776"-

'7

t

% ~

)>>

BIBLIOGRAPHY MODERN POWER PLANT ENGINEERINGi Wei sman 0 Eckart, 1985 Prentice-Hall Inc 4

~

1OCFP50. 10, "Requi r ement o f L.'ense. "

V ~

10CFR50.46, "Acceptance criteria for emergency core cooling systems for.'ight water nuclear power reactors."

10CFP50.55a, "Codes and Standards."

10CFR50.59)

"Changes, tests and experiments."

6.

10CFR50.70, Not i '.i cat ion". "

"Inspection,

Pecords, Pepor ts, 7.

10CFP50, Appendix A,

General Design Criterion 3-,

"Peact:".r coolant makeup."

8.

10CFP.

0, Appendi x A,

Gener al Desi gn Cr i ter i on 35,

" merqency core ccc ng."

10CFR50, Append x

A, General Design Criterion 36, "Inspection of emergency core cooing system."

10CFR50, Appendix A,

General Design Criterion 37, "Testing of Emergency Core Coo'ling systems. "

k V ~

>>50 iOCFR50)

Appendix B, I.'.

"Qua lty Assur ance Program"

10CFR50, Appendix 8, I II. "Desig" Ccntrol,"

iOCFR50, Appendix B, VI. "Document Control."

10CFR50, Appendix 8, X. "Inspections."

10CFP50, Appendi x 8, XI ~

"Test Con".rol. "

16.

10CFP50, Appendix B,

XIV.

"Inspection, Test and Operating Status."

17 10CFR50g Appendix B, XVI. "Corrective Action ~

18.

10CFRSO, Appendix 8, XVII. "Cua'lity Assurance Records."

19.

10CFR50g Appendix E,

F. ";raining.

20 ~

Federal

Reqister, Public Docket:

50-'" 0, Niagra Nohawk, Unit One, Nine Nile Point Therma.'uc!ear Peactor.

18

f

UNITED STATED OF AMERICA BEFOPE THE NUCLEAR REGULATORY COMMISSION AFFIDAVIT OF BEN L. RIDINGS I, Ben L. Ridings do make oath and say:

l.

My name is Ben L. Ridinqs.

I am a technical consultant for commercial nuclear power plants.

Over a span of some fifteen years, while working at some twenty four sites, I have specialized in reviewinq of licensing agreement (FSAR, Technical Specifications, Federal Codes and Regulations, ASME

Codes, etc.),

establishing administrative controls to meet these requirements and test programs to ensure compliance at all times.

My test programs and administrative

=ontrols established wh.'e under contract to various utilities are still in use today at many facilities.

2 ~

I have reviewed all of the relevant publicly available correspondence between the Nuclear Regulatory Commission and Niagra Mohawk during the relvant time span.

I am familar with NPC regulations and regulatory guidance qoverninq Hiqh Pressure Core Injection.

factual statement made in the attached Petition for Emergency Action and Request for public Hearing are true and correct to the best of my knowleqe and belief.

Ben L. Ridings Subscribed and sworn to before me this ~)~ day of (~

,1992.

My commisiom expires:

oi

I'h IL h

I yy

Part 50, 4PP. ~

Criterion 27-Combined rcactfuftp control systems capabftfty. The reactivity control systems shall be designed to have a corn.

blned capability, In confunction with poison addition by the emergency core cooling system.

of reliably controlling reactivity changes to assure that under postulated ac-cident conditions and with appropriate margin for stuck rods the capability to cool the core Ls maintained.

Criterion 2d-Rractfuftp ffmfta, The reac-tivitycontrol systems shall be designed with appropriate limits on the potential amount and rate of reactlvlty Increase to assure that the effects of postulated reactivity accidents can neither (1) result In damage to the reac-tor coolant pressure boundary greater than limited local yielding nor (2) sufficiently dis-turb the

core, Its support structures or other reactor pressure vessel Internals to impair significantly the capability to cool the core. These postulated reactivity acci-dents shall Include consideration of rod c]ection (unless prevented by positive means),

rod dropout, steam line rupture, changes In reactor coolant temperature and pressure. and cold water addition.

Criterion 2S-Protcctfon apafnst antfcf-patcd opcratfonaf occurrences.

Thc: protec-tion and reactivity control systems shall be designed to assure an extremely high proba-bility of accomplhhlng their safety func-tions In the event of anticipated operational occurrcnccL IV. F/ufd Systems Criterion 30-Quatfty of reactor coolant prcssure boundary. Components which are part of the reactor coolant pressure bounda-ry shall be designed, fabricated, erected, and tested to the highest quality standards prac-ticaL Means shall be provided for detecting and, to the extent practical, Identifying the location ot the source ot reactor coolanL leakage.

Crftcrfon 32-Fracture prcventfon of reac-tor coolant prcssure boundary. The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, tesL-lng, and postulated accident conditions (1)

Lhe boundary behaves In a

nonbrlttle manner and (2) the probability of rapidly Propagating tfacLurc Is minimized, Thc design shall reflect'onsideration of service temperatures and other conditions of the boundary material under operating. malnte.

nance, testing, and postulated accident con-ditions and the uncertainties ln determining (I) material properties, (2) the effects of Ir-radiation on material properties, (3) residu-a), steady state and transient stresses, and (4) shc ot flaws.

Crttcrion 32-Inspectfon of reactor cool-ant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed to permit (I) 546 10 CFR Ch. I (1-1%8 Edltfon) periodic inspection and testing of Important areas and features to assess their structural and leaktlght Integrity. and (2) an approplf.

ate material surveillance program for reactor pressure vessel.

Criterion 33-Reactor coolant makeup. h system to supply reactor coolant makeup for protection against small breaks In the reactor coolant pressure boundary shall be provided. The system safeLy function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rup-ture of small piping or other small compo.

nents which are part of the boundary. The system shall be designed to assure that for onslte electric power system operation (as.

sumlng otfslte power h not available) and for offsltc electric power system operation (assuming onslte power h not available) the system safety function can be accomplished using the piping, pumps. and valves used to maintain coolant Inventory during normal reactor operation.

Crftcrfon 36-Resfdual heat rcmooaf. h system to remove residual heat shall be pro.

vlded. The system safety function shall be to transfer fhslon product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy ln components and features, and suitable Interconnections, leak detection, and Isolation capabilities shall be provfded to assure that for onslte electric power system operation (assuming otfslte power h not available) and for offslte elec-tric power system operation (assuming onslte power h noL available) the system safety function can be accomplished, assum.

Ing a single failure.

Crftcrion JS-Emcrpcncy core coolfnp. h system to provide abundanL emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss ot reactor coolant at a tate such that (1) fuel and clad damage that could Interfere with continued effective core cooling ls prevented and (2) clad metal-water reaction Is limited to negli-gible amounts.

Suitable redundancy In components and feaLures, and suitable Interconnections. leak detection, hoist!on, and containment capa-bilities shall be provided to assure thaL for onslte electric power system operation (as-suming offsltc power Is not available) and for ottsltc electric Power system operation (assuming onslte power h not available) the system safety funcLlon can be acoomplhhed.

assuming a single failure.

Crftcrfon 36inspection of cmcrpcncp core cooHnp svstcm. The emergency core

~<

"i NucIddr Rd9ulotc 4

gzifng system sha appropriate period tant componcntso s

r

'sics. and PIPlng, Lo C pabIIIty of the sy Crftcrcon 37-'Fc cooifnp system. Th system shall be de.

>. cate Periodic Presst

~~ to assure (I) the st 7+tegrity of ILs comp "and performance

.of. the system. and

~'system as a whole

.dose to design as

.of,thc full operat!

the system Into c

ation of appllcab

.:Lion system, the

ind emergency pc

+ation of Lhe assoc

~

..."~Criterion Jd-C

~: h system to rem' containment shal

's'afety function s conshtent with L soclatcd systems.

and temperature

.int accident and

,'bly low lc,veh.

':'Suitable redur

'features, and sul I'detection, Isolatl bllltles shall be

':onslte electric p

>sumfng offslte t

, for offsfte elect:

'assulnlng onsltc

'- system safety fu assuming a slngl

".Criterion 39-

'eat removal sp.

removal system q~f appropriate pc>

tant component spray nozzles, l

@;c'. tegrlty and capi Critcnon 4()-

rcmovat system moval system appropriate pe al testing to a

~C>;

leaktlght intel thc operablllt active compon the operablllL) and under con as practical tl operaLlonal sec Into operation.

. g cable portions transfer betw 4

power sources.

soclated coolln Criterion C

cleanup.

Syst

~ 'h

.'l-l-88 Edition) rtlng of Important rs Lhelr structural id (2) an spproprl-program for the nntarrt makeup. A

'oolant makeup rs)I breaks In the boundary shall be ety function shall rd acceptable fuel ded as a result of

> leakage from the

)oundsry and rup.

Lher small compo-

.he boundary. The to assure Lhst for

em operation (as-not available) and

'ystem operation

not available) the sr be accomplished snd valves used to

)ry during normal

'eat rcmovaL A

1 heat shall be pro-

~ function shall be rct decay heat and the reactor core at

.ed acceptable fuel Lslgn conditions of sure boundary are n

nents and crcu.wcctlons, leak urpabllltles shall be for onslte electric r (assuming offslte md for offslte elec-

+ration (assuming diable) the system xompihhed, assum-rcy core cool(np. A snt emergency core L The system safety defer heat from the any loss of reactor rat (I) fuel and clad fere wlLh continued r prevented and (2) n h limited to negll-In components snd'terconnectlons, leak I containment caps-d Lo assure that for rstem operation (as-r noL available) and er system operation ts not available) the am be sccompl(shed.

e

~tion of emergency the emergency core Nuclear Regulatory Commission cooling system shall be designed to permit appropriate periodic InspecLlon of Impor-tant components. such as spray rings In the reactor pressure vessel, water Iniectlon nor ales. snd piping. to assure the Integrity and capability of the system.

Critcr(orr Jy-Test(rrp of emergency core coot(up system. The emergency core cooling sysLem shall be designed to permit appropri-ate periodic pressure: and functional testing to assure (L) thc structural and lesktlght In-tegrity of Its components. (2) the operablllty and perfonnance of the active components of the system. and (3) the operability of Lhe system ss ~ whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system Into operation, Including oper-ation of applicable portions of the protec-tion system. the transfer between normal and emergency power sources, and thc oper-ation of the associated cooling water system.

Criterion Jd-Corrta(rrmerrt heat rernovaL A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other as.

socfated systems. the containment pressure and temperature following any loss of.cool-ant accident and maintain them at accept-ably low levels.

Suitable redundancy ln components and features. and suitable Interconnections. leak detection. ho)ation. and containment capa-bilities shall be provided to assure that for onslte electric power system operation (as-suming of(site power Is noL available) snd g;

for offslte electric power system operation (assuming onslte power h not available) the system satety function can be accomplished, assuming s single failure.

Criferion JS-lrrspcct(orr of contafnmcnt heat removat system. The containment, heat removal system shall be designed to permit appropriate periodic Inspection of Impor-tant components,'uch as the torus, sumps, spray norx)es. and piping to assure the In-

. tegrity snd capability of the system.

Critcnon 40-Tert(ng ofconta(nmcnt heat rcmovat system. The containment heat re-moval system shall be designed to permit appropriate periodic pressure and function-al testing to assure (1) the structural and leaktlght Integr(Ly of Its components, (2) the operability, and performance of the active components of the system, and (3) the operability of the system as s whole.

snd under conditions as close to Lhe design as practical the performance ot the full operaLlonal sequence

(,hat brings the system Into operation. Including operation of appli-cable portions of the protection system, the transfer between normal and emergency power sources, snd the operation of Lhe as-sociated cooling water system.

Criterion 41-Con(a(rrmcrrt atmosphere cleanup.

Systems to control fhslon prod-5 Port 50, App. A

ucts, hydrogen, oxygen.

and other sub-stances which msy be released Into the reac-tor containment shall be provided as neces-sary to reduce, consistent with the function-ing of other associated systcrns. the concen-tration and quality of fission products re-leased to the environment following postu-lated accidents, and to control the concen-tration of hydrogen or oxygen and other substances In the containment atmosphere following postulated accidents to assure LhaL contslnmenL Inl.egrlty ls maintained.

Each system shall have suitable: redundan-cy ln components and features, and suitable Interconnections, leak detection, Isolation, snd containment capabilities to assure that for onslte electric power system operation (assuming offslte power h not available) snd for offslte electric power system operation (assuming onslte power h not available) its safety function can be accomplished, sssurn-Ing a single failure.

Criterion 42-Irrspcct(orr of containmcnt atmosphere cleanup systcmL The contain.

menL atmosphere cleanup systems shall be designed to permit appropriate periodic'In.

spectlon of Important components.

such as filter frames, ducts, and piping to assure the integrity and capability of the systems.

Criterion 43-Testing of corrtafnmcrrt at-mosphere cleanup systems The containment atmosphere cleanup systems shall be de-signed to permit appropriate periodic pres-sure snd functional testing to assure (I) the structural snd leaktlght Integrity of Its com-

ponents, (2) the operability and perform-ance of the active components ot the sys-tems such as fans. filters, dampers, pumps, snd valves and (3) the operability of the sys-tems as a whole and, under conditions ss close to design as practical, the performance of the full operational sequence that brings the systems Into operation, Including oper-ation of applicable portions of the protec-tion system, the transfer between normal and emergency power sources, and the oper-ation of associated systems.

Critcriou dd-Coot(ng roatcr. A system to transfer heat from structures, systems. and components Important to safety, to an ulti-mate heat sink shall be provide* The system safety function shall be to transfer the combined heat load ot these structures, sysLems, and components under normal op.

crating and accident conditions.

Suitable redundancy In components and features, snd suitable Interconnections, leak deLectlon, and Isolation capabilities shall be provided to assure thaL for onslte electric power system operation (assuming offslte power Is not avallablc) and for offslte elec-tric power system operation (assuming onslte power h not available) the system safety function can be sccomplhhed. assum-ing a single failure.

0

LIHITIHG CONDITION FOR OPERATION SURVEILLANCE AEQUI AEHL'NT 3.1.8 IIIITI POESSUOE COOLANT II~JfCTION 4.1.0 IIIG)I PAESSUAE COOLANT IHJECT IOH Applies to the operational status of the high pressure coolant injection system.

~0b EaCti Ve:

To assure the capability of the high pressure coolant injection system to cool reactor fuel in the event of a loss-of-coolant accident.

Applies to the periodic testi>>g )equiruiiiunts for the high pressure conla>>t i>>juct1>>>> systul>>.

~Ob ective:

To verify the operability of the high )Tressure coolant injection system.

~if)

)

a'i

~~

a.

Ouring the power operating con-dition whenever the reactor coolant pressure is greater than 110 psig and the reactor coolant temperature greater than saturation telllperature, the high pressure coolant 1njection system shall be operable except as speci fied in Speci fica tion "b" below.

b.

I f a redu>>dant component of the high pressure coolant injection sys tern becomes inoperable the high pressure coolant 1njection shall be considered operable proviiied that the component is returned t,o an operable condi tion within 15 days and the additional sur-veillance required is performed.

~Seci f1 cation:

The high pressure coolant 1>>jectio>> sur-veillance shall be performed as indicated below:.

a.

At leaSt PnleT eU".O eeatlntirCTCInni ss'"

~ r-waaaec,~

Automatic start.-up of the hii)h pressure coolant 1njection system shall be demo>>-

s tra ted.

b.

A Pump operability shall be duton>>I>>ud.

71

g I

LIH r

C S~UAVE LLAQCE A~El)i AEML'O'I'.

If Specification "a" and "b" are not

met, a normal orderly shutdown shall be initiated w> thin one hour and reactor coolant pressure and temp-erature shall be reduced to less than 110 psig and saturation temperature within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

Surveillance wi th I~no crab le Coiiginnunt When a component becomes inoperabio Its redundant component shall hu deuens tratud to be operable iiwwdiately and daily thereafter.

1'

F 8ASES FOR 3.1.8 AND 4 ~ 1.8 HIGH PRESSURE COOLANT IHJK(.) )OM a High Pressure Coolant In]ect)on System (HPCI) is provided to ensure adequate core cooling in the unlikely event of a)) reactor coolant l)'ne'reak.

.The HPCI

.System.)s required for line breaks ~hich exceed the capability of the ntrol Rod Drive pumps and which are not large enough to allow 'fast'enough depressurizat)on for core spray to be fective.

e set of high pressure coolant )n]ect)on pumps cons)sts of a condensate

pump, a feedwater booster pump and a motor iven feedwater pump.

One set of pumps is capable of delivering 3,800 gpm to the reactor vessel at reactor essure.

The performance capability of HPCI alone and in con]unct)on with other systems to provide adequate core oling for a spectrum of line breaks is discussed in the Fifth Supplement of the FSAR.

I dete'ning the operability of th~ HPCI System, the required performance capability of various components sha11 be ins)dered.

~ The HPCI System shall be capable of meeting its pump head versus flow curve.

.p The motor driven feedwater pump shall be capable of automatic initiation upon receipt of either an automatic turbine trip signal or reactor

)ow-water-.level signal.

.~ The Condenser hotwell

~eve) shall not be less than 51 inches (75,000 gallons).

The 'Condensate storage tanks inventory shall not be less than 105,000 gallons.

=

The motor-driven feedwater pump will automatically trip if reactor high water level is sustained for ten seconds and the associated pump downstream flow control valve and low flow control valve are not closed.

jring reactor start-up, operation and

shutdown, the condensate and feedwater booster pumps are in operation, At aactor pressures up to 450 ps)g, these pumps are capable of supplying the required 3,800 gpm.

Above 450 psig a

ator-dr)ven-feedwater pump is necessary tu provide the required flow rate.

he capability of the condensate, feedwa'ooster and motor driven feedwater pumps will be demonstrated by their peration as part of the feedwater supply during normal station operat)on.

Stand-by pumps wil1 be placed in service t least quarter)y to supply feedwater during station operation.

An automat)c system initiation test will be erformed at least once per operating cycle.

This will involve adtomatic starting of the motor driven feedwater pumps nd flow to the reactor vessel.

Revised October I, 1906

.l

~

l

I.

HIGH-PRESSURE COOLANT INJECTION 1.0 Desi n Bases The high-pressure coolant injection (HPCI) system is an operating mode of the feedwater system available in the event of a small reactor coolant line break which exceeds the capability of the control rod drive pumps (0.003 ft2).

HPCI along with one emergency cooling system has the capability of keeping the swollen reactor coolant level above the top of active fuel for small reactor coolant boundary breaks up to 0.07 ft2 for at least 1000 seconds.

The HPCI system with one of the two emergency cooling systems and two core spray systems, will provide core cooling for the complete spectrum of break sizes up to the maximum design basis recirculation discharge line break (5.446 ft2).

Its primary purpose is to:

Rev.

7 a.

provide adequate cooling of the reactor core under abnormal and accident conditions.

b.

remove the heat from radioactive decay and residual heat from the reactor core at such a

rate that fuel clad melting would be prevented.

c.

provide for continuity of core cooling over the complete range of postulated break sizes in the primary system process barrier.

HPCI is not an engineered safeguards system and is not considered in any Loss of Coolant Accident Analyses.

It is discussed in this section because of its capability to provide makeup water at reactor oper'ating pressure.

Rev.

7 2.0 S stem Desi n

The HPCI system utilizes the two condensate storage

tanks, the main condenser hotwell, two condensate

pumps, condensate demineralizers, two feedwater booster

pumps, feedwater

heaters, two motor-driven feedwater

pumps, an integrated control system and all associated piping and valves.

The system is capable of delivering 7600 gpm into the reactor vessel at reactor pressure when using two trains of feedwater pumps.

The condensate and feedwater booster pumps are capable of supplying the required 3,800 gpm at approximately reactor pressures up to 270 psig.

Above 270 psig a motor-driven feedwater pump is necessary to provide the required flow rate.

Rev.

7 Rev.

7

VII-61a The feedwater system pumps have recirculation lines w1th air operated flow control valves to prevent the'umps from operating against a closed system.

In the event of loss of air pressure, these valves open recycling part of the HPCI flow to the hotwell.

HPCI flow would be reduced to approximately 3,000 gpm at a

reactor pressure of 1,150 psig and 3,800 gpm at a reactor pressure of 940 psig.

Condensate inventory is maintained at an available minimum volume of 180,000 gallons.

Rev.

7 3.0 Desi n Evaluation During a loss-of-coolant accident within the drywell, high drywell pressure due to a line break will cause a reactor scram.

This automatic scram will cause a

turbine trip after a five-second delay.

In order to prevent cladding temperatures from exceeding their maximum limit for the entire spectrum of breaks, the 3800 gpm (from one train of HPCI/feedwater pumps) would have to be available immediately.

Feedwater flow would be available for considerable time from the shaft-driven feedwater pump.

The shaft-driven feedwater pump would coast down while the electric motor-driven condensate pumps and feedwater booster pumps would continue to operate.

The coast down time to reach 3,800 gpm delivery to the core is approximately 3.2 minutes (Figure VII-17), since both the condensate and feedwater booster pumps will continue to operate on off-site power.

The turbine trip will signal the motor-driven feedwater pump to start.

The signal will be simultaneous with the start of the shaft pump coast down.

The motor-driven feedwater pump will be up to'speed and capable of supplying 3,800 gpm in about ten seconds.

As a backup, low reactor water level will also signal the motor-driven pump to start.

The initiation signal transfers control from the normal feedwater to the HPCI instrumentation and controller which has been continuously tracking the normal feedwater control signal.

Thus there will be a

continuous supply of feedwater to the reactor.

The HPCI single element control system will attempt to maintain reactor vessel water level at 65 inches or 72 inches (depending upon which pump, 11 or 12 respectively, is in service) with a maximum feedwater flow limit of 3800 gpm.

Rev.

7

'(t7

~

~

I

A sustained high reactor water level reactor protection system signal coincident with an open feedwater flow control valve will selectively trip the associated feedwater pump.

The clutch of the shaft-driven pump will also be disengaged immediately upon high reactor water level.

Should the reactor water level reach the low level scram setpoint the motor driven pump that tripped on high reactor ~ater level will restart.

Necessary feedwater pump recirculation is provided to allo~ for continued pump operation with the flow control valve closed.

As feedwater is pumped out of the condenser

hotwell, through the selected equipment of the condensate and feedwater systems and into the reactor, the condenser hotwell level will fall.

Since condensed steam from the turbine no longer replenishes the condenser hotwell, condensate will be transferred from the condensate storage tanks to the hotwell for makeup.

The feedwater system pumps operate on 4160 v.

Hhen the plant is in operation, the power is supplied from the main generator through the station service transformer when the generator is on-line and connected to the grid.

Hhen the main generator is off-line, the feedwater pumps are supplied with normal off-site power from the 115 KV 'system through the reserve transformers.

If a HPCI initiation signal should occur, all HPCI/feedwater system pumps would start immediately with two feedwater pump trains available for HPCI injection using the single element feedwater control system for reactor vessel level control. If a major power disturbance were to occur that resulted in loss of the 115 KV power supply to the Nine Nile Point 115 KV bus, power would be restored from a generator located at the Bennetts Bridge Hydro Station.

This generator would have the capacity of supplying approximately 6,000 KVA which is sufficient to operate one train of HPCI/feedwater system pumps.

If HPCI initiation were to occur, the preferred feedwater train pumps (feedwater pump 12, feedwater booster pump 13, condensate pump 13) would start.

The non-preferred train pumps would be locked out on loss of off-site power and not start until the operator manually reset the lock out.

If a preferred train pump had been locked out prior to the loss of off-site power, it would remain locked out and the non-preferred train backup pump would automatically start on HPCI initiation. If both the preferred and backup pumps are running, the preferred pump would remain in service and the backup pump will trip.

The

p ~

Q

~

0'

VII-62a use of a Bennetts Bridge hydro generator, while not equivalent to an on-site emergency power source, provides a highly reliable alternate off-site power supply for the HPCI function of the feedwater system.

4.0 Tests and Ins ections ev.

7 Tests and inspections of the various components are described in Section XI - Steam to Power Conversion.

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<a Mr. B. Ralph Sylvia November 19, 1992 This requirement affects one respondent and, therefore, is not subject to Office of Management and Budget review under P.L.96-511.

Sincerely, Original Signed By:

Enclosure:

10 CFR 2.206 Petition dated October 17,, 1992 from Ben L. Ridings',,'

cc w/enclosure:

See next page I

1J 1

DISTRIBUTION:

,Docket File

.NRC

& Local PDRs PDI-1 Reading SVarga JCalvo RACapra CVogan DBrinkman OGC ACRS (10)

CCowgill, RGN-1 Plant file Donald S.

Brinkman, Senior Project Manager Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation OFFICE PDI-1: LA PDI-1: PM OGC PDI-1: D CVo an H DBrinkman:avl JGol b

g RACa ra DATE

'l4 15/92 92 92 Ii l 92 OFFICIAL RECORD COPY

.fILENAME: NM184890. LTR

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