DD-96-20, Responds to Petition Filed on 960120 on Behalf of Petitioner Re Plant Operated by Licensee.Nrr Denied Licensee Request for Immediate Action Suspending Operating License.Director'S Decision DD-96-20 & Notice Issuing Director'S Decision Encl

Responds to Petition Filed on 960120 on Behalf of Petitioner Re Plant Operated by Licensee.Nrr Denied Licensee Request for Immediate Action Suspending Operating License.Director'S Decision DD-96-20 & Notice Issuing Director'S Decision Encl

ML20134M123
Person / Time
Site:	Maine Yankee
Issue date:	11/20/1996
From:	Miraglia F NRC (Affiliation Not Assigned)
To:	Burt A AFFILIATION NOT ASSIGNED
References
CON-#496-18069 2.206, DD-96-20, NUDOCS 9611220232
Download: ML20134M123 (13)

Text

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jmg & UNITED STATES s o NUCLEAR REGULATORY COMMISSION DOCKEIED o WASHINGTON, D.C. 2066H001 November 20, 1996 b*****# '96 NOV 20 P4 37 Ms. Anne D. Burt Corresponding Secretary Of FIM OF SE SE f/6Y Friends of the Coast - Opposing Nuclear Pollution 00CKE TmG & 3EWiCf.

Post Office Box 98 r;R!&

Edgecomb, Maine 04556

Dear Ms. Burt:

I am responding to the petition you filed on January 20, 1996 (the Petition), I on behalf of the Friends of the Coast - Opposing Nuclear Pollution (Petitioner), in regard to the Maine Yankee Atomic Power Station (Maine Yankee), operated by the Maine Yankee Atomic Power. Company (the licensee). J Your petition was considered pursuant to Title 10 of the Code of Federal Reaulations, Section 2.206 (10 CFR 2.206). The Petition requests that the ,

Commission take expedited action to (1) suspend the operating license of Maine Yankee pending resolution of the Petition; (2) examine and test by plug sampling - or other methods approved by the American Society of Mechanical Engineers - all large piping welds that may have been susceptible to micro-fissures at the time of construction; (3) reanalyze the Maine Yankee containment as one located in an area where seismic risk is not " low"; (4) reduce the licensed operating capacity of Maine Yankee to a level consistent with a flawed containment and/or flawed reactor coolant piping welds; (5) hold 1 an informal public hearing in the area of the plant regarding the Petition; )

and (6) place the Petitioner on service and mailing lists relevant to the group's interests in safety at Maine Yankee and intention to participate in all public forums opened by the Nuclear Regulatory Commission (NRC).

By letter dated May 13, 1996, Mr. William Russell, Director, Office of Nuclear Reactor Regulation, acknowledged the NRC's receipt of your Petition, and, for i the reasons stated in the letter, denied your request for immediate action '

suspending the operating license or reducing the licensed operating capacity of Maine Yankee (Requests 1 and, in part, 4.). In addition, for reasons  !

stated in the May 13, 1996, letter, Mr. Russell denied your request for an informal hearing (Request 5). Mr. Russell also stated in the May 13, 1996, letter that your request that the NRC place you on service and mailing lists relevant to your interests in safety at Maine Yankee and your intention to participate in all public forums opened by the NRC (Request 6) was moot, as your attorney had already been added to the Maine Yankee service list.

The remaining specific issues that you raised that were the basis for Requests 2, 3, and 4 of your Petition dated January 20, 1996 (identified above), are fully addressed in the enclosed Director Decision (DD-96-20).

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Ms. /c  : C ert  :

For the reasons given in the enclosed Director's Decision under 10 CFR 2.206, l your remaining requests for NRC action have been denied. A copy of the l decision will be filed with the Secretary of the Commission for the i Commission's review in accordance with 10 CFR 2.206(c). As provided by this I regulation, the decision will constitute the final action of.the Commission 25 days after the date of issuance of the decision unless the Commission, on its own motion, institutes a review of the decision within that time.  !

I have also enclosed a copy of the notice of " Issuance of Director's Decision Under 10 CFR 2.206." This notice includes the complete text of DD-96-20 and is being filed with the Office of the Federal Register for publication.

Sincerely, ,  !

l AW W FrankJbirag Acting Director .

Office of Nuclear Reactor Regulation  !

Docket No. 50-309 ( 3 2.0 6)

Enclosures:

1. Director's Decision DD-96-20 ,

2. Notice l cc w/ enclosures:

See next page

Anne D. Burt cc:

Mr. Charles 8. Brinkman Mr. Robert W. Blackmore Manager - Washington Nuclear Plant Manager Operations Maine Yankee Atomic Power Station AB8 Combustion Engineering P.O. Box 408 12300 Twinbrook Parkway, Suite 330 Wiscasset, ME 04578 Rockville, MD 20852

Mr. G. D. Whittier, Vice President

, Thomas G. Dignan, Jr., Esquire Licensing and Engineering Ropes & Gray Maine Yankee Atomic Power Company 4

One International Place 329 Bath Road i Boston, MA 02110-2624 Brunswick, ME 04011 l

! l l

Mr. Uldis Vanags Mr. Patrick J. Dostie

, State Nuclear Safety Advisor State of Maine Nuclear Safety i State Planning Office Inspector I State House Station #38 Maine Yankee Atomic Power Station I

Augusta, ME 04333 P.O. Box 408 l Wiscasset, ME 04578 l

] Mr. P. L. Anderson, Project Manager I j Yankee Atomic Electric Company Mr. Graham M. Leitch 580 Main Street Vice President, Operations l Bolton, MA 01740-1398 Maine Yankee Atomic Power Station

P.O. Box 408  ;

Regional Administrator, Region 1 Wiscasset, ME 04578 l l U.S. Nuclear Regulatory Commission 4 475 Allendale Road Mary Ann Lynch, Esquirc King of Prussia, PA 19406 Maine Yankee Atomic Power Company )

329 Bath Road First Selectman of Wiscasset Brunswick, ME 04578 ,

Municipal Building U.S. Route 1 Mr. Jonathan M. Block Wiscasset, ME 04578 Attorney at Law P.O. Box 566 Mr. J. T. Yeroken Putney, VT 05346-0566 Senior Resident Inspector Maine Yankee Atomic Power Station Mr. Charles D. Frizzle, President U.S. Nuclear Regulatory Commission Maine Yankee Atomic Power Company P.O. Box E 329 Bath Road Wiscasset, ME 04578 Brunswick, ME 04011 Mr. James R. Hebert, Manager Nuclear Engineering and Licensing Maine Yankee Atomic Power Company 329 Bath Road Brunswick, ME 04011 I

1 i

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FKIENDS Post Office Box of 98, the COAST Edgecomb, Maine -04556OPPOSINdQtif$hR POLLllTION hk6ne/ fax - 207-882 - 6000

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'96 NDY 20 P4 :5}anuary20,1996 William T. Russell, Director Office of Nuclear Reactor Regulation 0FFICE CF SECRE TARY l

United States Nuclear Regulatory Commission DOCKE[h ? SERVICE Washington, DC 20555 - 0001 M'M i

Dear Mr. Russell,

i i Pursuant to the provisions of 10 CFR 2.206, Friende of the Ca==* - Onna ine Wclene j Polhitian- a non-profit organization incorporated in the State of Maine, petitions prompt and 1

thorough consideration of the following unresolved or insufficiently considered safety-related

issues pertaining to Maine Yankee Nuclear Power Station ( MYAPS )

i 1. Containment is inadequate for power operation in excess of original license and may be i inadequate for original power operation limits based on insupportable original design

acceptance criteria. The containment at MYAPS was designed and constructed without 2

diagonal reinforcement rod. Upon our best information and belief, the Atomic Energy ,

j Commission staff recommended to the commission that a license amendment permitting this j type of construction be allowed, '...for this plant and this plant only due to low seismic risk.'

Early in 1979 the MYAPS was shaken by an earthquake of 4.2 magnitude and epicentered 4 less than ten miles from plant site. The NRC then ordered the shutdown of five nuclear power

stations including MYAPS until piping and piping supports could be seismicly qualified.

j Upon our best information and belief, there is no public record that NRC did a second i evaluation of MYAPS marginally acceptable containment design, nor can we find any record j of reevaluation prior to any subsequent granting oflicense amendments to operate at increased j power. Enclosed are sample pages from 1%8 and 1971 MYAPS/ AEC correspondence files

indicating the situation. It is petitioner's belief that MYAPS unique contamment design is first j mentioned in construction license amendments one and two. Complete files are, of course, at

your disposal j .

2. MYAPS Emergency Core Cooling System, Primary Coolant Piping, and other larg:

i piping has not be adequately analyzed for materials degradation to ensure integrity at power i operation in excess of original license limits or under accident conditions. Such analysis could i prove that piping integrity is inadequate for continued operation at original limits. A review of l MYAPS construction license amendment 30 will show that the Atomic Energy Commission j was concerned enough with the appearance of" micro-fissures " in reactor coolant system welds to appoint a " task force". In 1971 the AEC's concem prompted studies and reports by

, Battelle Columbus Laboratories, Stone and Webster Engineering, and consultant Dr. Ernest j Nipes of Renssalaer Polytechnic Institute which generally concluded that the micro-fissures i would not propagate or grow under foreseeable conditions.

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{ However, petitioners maintain concem for the micro-fissures was dismissed prior to a 3 heightened awareness of embrittlement phenomena in reactor s e.;sel walls and welds. Of l 2

particular concern to petitioners are those large pipe welds attaching to, or next to, the reactor j

vessel which have endured 23 years of corrosion, stress, vibration, and radiation and which

! may fail initiating a Loss of Coolant Accident or which may be subject to thermal shock failure initiated by use of the ECCS.

i

_. )

Enclosed are sample pages from the AEC/ MYAPS 1971 correspondence file which i supply references to the issue microtissures in welds on the MYAPS reactor coolant system.

j Pursuant tothe conditions of 10 CFR 2.206, petitioners request the following actiorts be undertaken by b'RC:

1. Suspen.' the operating licen.,e of .daine Y ee Atomic Power Station until the

! above issues are thoroughly examined and resolv xamine and test by plug sampling or 2

other ASME approved methods all large piping welds which may have been susceptible to micro-fissures at the time of construction. Reanalyze MYAPS containment as one located in an area where seismic risk is not " low". S '

i

! Q I. Failing suspension oflicense, or should license be restored following examination '

of the above issues, reduce the license operating capacity of MYAPS to levels consistent with i

a flawed containment and or flawed reactor coolant piping welds.

f 1 Provide an infont-' public hearing regarding this petition in the plant area.

J j 4. Act on this petition in an expedited manner due to what the petitioners believe are

! the serious implications of the concems raised.

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,y $. Place Friends of the Coast on such service and/or mailing lists as may.be relevant to  :

ourinterest in safety at MYAPS and our intention to participate in all public forums opened

! by NRC, ,

Thank you for your Attention. Please address all correspondence to

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{ Anne D. Burt Corresponding Secretary, Friends of the Coast - Opposing Nuclear Pollution l Post Office Box 98, Edgecomb, Maine 04556 i

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w United States Atc..ic m Energ/ Co.iticnion L'asitin'; ton, D. C. 0D.765 Attention: Director, Divicien ef T.cactor Licensing Daar Sir:

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!!olicy cal ::1.i;3 c.ited .kcc:.cr b,.,ici 6 hoar For:cs *.r.co isted 1ith h h- Jr.

Provi:1or. fer ?.euistrac t o ." .-

quako Loadint' docu-ents the 1. ..i.rlal prear.nt:d by Irefcasor R/ic J at a meetin; trith the tito:deTlic rce.ON is avb-utted to rv.b.:t. ntiato deci".in'; the on Decenber lit, 1967.

reactor contain:. ant v.:ing ody c.1et:rtfcrential and vertical rcbara to provido tho meritbrano strcncth.

Respectfully sulnitted, luI;C Yl.TT'!2 ATO:IIC M.T4R CO::F/JTf e-> *> - - - - >

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1 Chspiun /.vecun Rockville, Maryland 20!)S2 Centlemen:

In accordance with ths ur.dcrutendings rc chsd Augese 1 , 1961, botvsen the Ato.ic Cncegy Ccusio: ion and the Department of Hamith, Educatica, cro the following docu:nento: and (felfeira, attact::d for your information "AT':: v'J:-~; i.:;mi:: n -r; c,(;

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/nendennt Nos.1 ct:d 2 datM .knicry 15, 1000 to

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dated Decccber 2), J.9C( r.*:n F. v:en, tally and

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i Provisions for l'.anJ::tnnce t.o 'it t.9ntial Lhear i Forces Accociated with Du t.bm s .o Lundire. - Con. '

taintent Chell.of Psinc YurAcc belcar Pt,wr Plant. "

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l 5.1 '.:hy do you s t.it e ( p.2 3e 5. ~, ( a ) et the rs.t!:) that Act spee<a t Publication 17 uilt be used in the design of the prir.ary shicid l ,

wall where the Act 313-t.,3 Code is not applicabic, since ACI t

l Special Publicat.ivn L7 L.s a detailcJ c::pansion of the ACI i 318-63 Code?

I

' 5.2 Indicate the :t reu vad/or str.~.ia criteria that have been applied to the desir.n of the large openings in the containment and the nothods used La ve.'t- *!-- adcetuacy of these laruc l t. .s eE e-openings. r... ,,, . .

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.<,o .,n t.,- *u4v l 5.3 ror the ir.terior s t ruc t.o r.w . indicate cbc bases for the loads j given in f ee t i. n 5.1. " . / . I ..f t he l'S.\:t an 1 1.hc load combinations .

s t re s s an l/ o r ..t ra in c r i t.. . i .. . .ied authod ot' desir,n.

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which stre sec .:rc critica! ..nd > Late their level ti. i 5.4 Describe the desiv,n of the radial shcar ladders (dencribed on pago 5-20 of the FSAn) in the containment vall, including g

- the extent of ecacrete presui.ed to act with.these ladders.

i f5.5 Since no diagonal reinforcing has been provided in the contain-ment wall for scismic shears, indicate, for comparison, the l { maximum scisnic shear stress ir the concrete and ILnce, and l

• the nonnal and chcar stresses in the radial reinforcing assuming (a) that all the load is carricI by the concrete without liner participarlon, and (b) that all the load is carried by the L liner.

5.6 speci fy and describe the des t y.r of the neutron shield tank j

including criteria, codes, dest n rethods, and applicable quality control provisions. include su (icient details of the reactor l

supports on this tank to permit an evaluation of the design l adequacy.

j.

5.7 Define the frequency with uhich 12 inch Ic,gths of reinforcing bars have been tested by an ind pendout laboratory as a check on the acceptability of the hca s of 50,000 psi yictd point steel used in construction. Al o, state uhother Cull size i

samples have been tested to ver fy compliance with the design j strengths.

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Describe where the l'repakt retiu 1 of placing concretc uns used j 5.8 and the chechs that have bcen rzle on the quality and penceration of in-situ Propakt concrete, ine'.uding the results of these checks.

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5.9 :escribe any deviatinn, a

for the 11. f ros s;.ec i t ica d i.mn.;jona l c e r e, in,;c :

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required, ao 1. ..

ether th- c r.:e s.rr;ctive ncasurcs tahon sehere t i..!

dinensiona l criteria. liner meetu the required 5.10 j Stato uhether are ..ciding has been useil on reinforcint; s bar for cicher stre: Nth or tetch velds. anJ list the acecptance criteria used for such we lding. ..'

used for bo iJ anJ ancha ta.;c ot~ o, describe the criteria that is in rension reinio ciny, hari in concrete .

discontinuity eones).(such as thn concrc c in the dree and l 5.11 j $000 psi concrete was placed, ina:hi required, Describe the quality c

hly st.ren

,ed t

j areas of the containucnr. instead ot' the 3,000 psi concretc used in other area ..

5.12 Describe the pemanent available to record contalumentin-place in.irrumentation which will be accident pressures and t emperaturcs. operating, accident and post-5.13 State whccher the containment to the calculated peak accidenthaspres:;ure been designed at to permit testing lifetime. anytime during plant 5.14 State whether the containment integrated leak rate test uilt be fully perforrned installed. with the penetrations and associa cd pipiu; systems for this test, If the penetrations are to b 1,lanked off by comparison with conditions which may prevaljustify the validity ot conduct completion of installation of the piping syste S.follouing i.15 ruscribe the seismograph installation that wili be used, the location the criteria with and maintenance respect program for the installationinetuding to assesr.ments to be made and plant

. State operatinn to be pernitted in the event

! of instrument readings L in theEarthquake.

Basis range of the Operating tlasts Earthquake and the Design i

.16 Provide a dir.cussion of the performance charac aristics of protectiveaccident uithstand conting and paints used within the :ontain::icnt to conditions, includin:: consii2 ration of spray var.hdown, steau environment, and jet impi co,emci t e f fec t s. Also, include' an evaluation of t'53 potential imp 3Lrut it of the performance capabilities fouling of heat of ene,incerud sa fety features, due to flow blockage transfer surfaces, or other eve its that mi t;h t ,

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result from failure ot the protective cuattnrs and paints.

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l Septe:ber 21, 1971

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Attention: i $6Fg g 7*. ,/h Director, Division of Reactor Li 13 2 .. .

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Dear Sirs:

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4 dated SeptemberAMEND.5NT No. 30 to LICENSE APPLIC i

26. 1967 ATION i (Oceket No. 50-309) j the Cot =1ssion's Rules and RPursuant to the Atocic Energy Ac '

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ATOMIC PO'. R COMPANY hereby aegulations issued , and thereundert of 1 the attached information ends its license application by s bMAINE YANKEE i , 1. u mitting t

Battelle Columbus LaboratoriesFifty or Maine Yankee by the copies of a -

I

Reactor Coolant Systen" a dated Sof Type 316 Sta epte ber roc Maine Yankee

{ 2. 17, 1971.

t Fifty copies of a report Rensselaer Polytechnic Institutby,Dr. Ernest F. Nippes of -

dated 9/21/71. Stainless Steel Weld:ents at Maine entitled " Anal i

3. e Yankee Atomic Power Station"

\ Fifty copies of Stone & Webster E j MYS-!.819 dated September 4 14, 1971ngineering Corporation letter 1

i Eighty copies of revised FSAR pag

5. e 13-6 dated 9/71.

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i In response to questions asked by th i,

pipe and vessel insulation is submitt dreactor co e: ng information in 1

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i Septenter 21,197i T.s.,6 l <- ;.. ' sns tre essa

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v ainc Yankaa Atomic Potter Compcny Q f[U.! Y Engineerlag Offico f!Lt;i I i i.U Y i

Turnpike Read (acute 9) 6] SEP2 219712 3 Vestboro, P. ass. 015ol 8 L C V:t c.v.t g #f,I.S[j' Attention: sa.: acas q

.  ::r. 3. B. Beck 1cy, Project Engineer

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Subjec t  :

j Analysis of Stain 1 css Steal Vcids:nt: at

aine Yankee Atomic Power Station Re fe rence s:

(1) Final Report

} Steet Ucid :'acc' "1.aboratory Studies of Type 316 Stainless ,

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• Battelle Col'.m. bus Lai: oratories. 17, Septemberfrem 1971 Maine' Yanke

{ (2) Letter from ::. R. Gilbert, i Project Sgi:.eer, Stone and i Webster Engineering Corporation September 14, 1971.

, Project to B. 3. BeckleyE -

Subject:

Yankee Atomic Power Station." '".tierofissured Uelds Maine

Dear Mr. Beckley:

The following su=arizes cy cnalysis of the Type 316 stainl i in the reactor piping systems of tha !!aine Yanhea Ate iess st.cel walde.cnts j r c Potter Stations i

_BACXCCOU?!D,

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Weld stects, and metal other microfissures alloys. have been discovered y of steels,in astainicss variet are usually found only by matallo;;raphic technieues ,

re short in because th len;;th, gancrally in the rants from 2 to 80 etts, and very narro r ..

Microfi.tsures in austenitic stainless steels are scoccal i

have been rehested by a subsequent weld pass. oods which involve a hot-tearing mechanis in t;hich contraction stressesBus, micecEissuring is thoug grain boundaries which have been tirated by reheating, rupture austenitic i

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occurs thesefe r ri te wa s ~se order of 11 t"o service y g, 3 of th e is of tempe ra tu re s .in order to

{ less than 1100, upper limit ten specified As statutae signa. .

F.

i can be set highce, if the sa to be in thea result of enginee re ge of The eechanism by which ferrite rvice temperature is i . at present.

i

! energy of an aurtentta-farritea ebou bounda ry. r. dOne of the more rece \

i be wetvet by the liquid, : harassThus, when involves the lower liquation occursn interfaci a ary c lo

i. boundary ,

an aus:enite-ferritean austent:2-sus:cnite boundar boundary

! and therefore ruptures, forming an iby liquidwould isbe.

not not wouldabic t A

i ntergranular separation.n the imposed c Although the occurrance of microfi ses i

! for many years, the effect of et

! been extensively ra;*orted inerofissures th on mechanical proper

• button has been re:c e open literature. recognized es has not I ontests In the perfor-fatigue*ds:rcn;nt1;. made by at ;h of nickel-base alloys Yeniscavich rted who repoThe cost impo i

I with effectivo diaN:ers up to 70 roe. e: pe.rature gontainin; and 550 r, he concluded chramiu a and iron.t

} fatigue life in larp components cils would have no =casurable as 70 per square in:S had no meas. In rddition,

e f fec t on sures connect. did not propegate during urablefati gue effect on fetipe life andlocal fissur i

The effcc: ,

!) basis that thein low-cycle than in high-cyciof fissures in olo;;ering inter- th2 te t duces the stress repea:cd plastic concentration s

defornation accompc fa:igue; th a ned on the i Examination of f atigue samples af associated with defects in ductileanying low-cyc

! had initiated fatigue cracks -

ter failure showed that ough

, others had not, and that alth some fissures ma te ria ls.

!: cracks was not a function of fissur e site.

3 NAIN YA'lKEE -

the initiation of ' fatigue i

The reactor coolant system large di steel piping. Af ter heat -

1 l Type 'l16 stainless steel, treat with~.enta f of the carbon secelameter piping carbon is stai to this piping for subsequent errite concent in excess of 57 piping, safe ends of site welding toarecomponents, other Type 316 , were staijoined l which con i

i with Type 316 stainless steel of 107 ferrite. n ess steel i ents, i. e. , pipe -to-valve, pipe -tothose welds involving stainlessThe on 15 welds, ninc of which were conpl e:ed willwith b low-ferricepump, and valve-:o-pump I

! maining six welds have been or electrodes, tatning a minimum of 57. ferrit iho re -

i evaluated in this analysis.

e and which therefore will noton-be con i s ered or dd b (n&m4/ NO .

i 1

l DOCKETED USNRC 00-96-20 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION '96 Nfr! 20 P4 :57 0FFICE OF NUCLEAR REACTOR REGULATION Frank J. Miraglia, Acting DirehkT (

In the Matter of )

)

MAINE YANKEE ATOMIC POWER COMPANY ) Docket No. 50-309 l

)

(Maine Yankee Atomic Power Station) ) (10 CFR 2.206)

) I i

! DIRECTOR'S DECISION UNDER 10 CFR 2.206 I. INTRODUCTION 4

By letter dated January 20, 1996, Ms. Anne D. Burt filed a Petition with the U.S. Nuclear Regulatory Commission (NRC), pursuant to 10 CFR 2.206, on behalf of the Friends of the Coast - Opposing Nuclear Pollution (the Petitioner) requesting that actions be taken regarding the Maine Yankee Atomic Power Station (Maine Yankee), operated by the Maine Yankee Atomic Power Company (the licensee). The Petition requests that the Commission take expedited action to (1) suspend the operating license of Maine Yankee pending resolution of the Petition; (2) examine and test by plug sampling - or other methods approved by the American Society of Mechanical Engineers - all large piping welds that may have been susceptible to micro-fissures at the time of construction; (3) reanalyze the Maine Yankee containment as one located in an area where seismic risk is not " low"; (4) reduce the licensed operating capacity of Maine Yankee to a level consistent with a flawed containment and/or flawed reactor coolant piping welds; (5) hold an informal public hearing in the area of the plant regarding the Petition; and (6) place the Petitioner on service and mailing lists relevant to the group's interests in safety at Maine Yankee and intention to participate in all public forums opened by the NRC.

%t22e>W ng.

._ _ _ = . - - - . . = . - --- -. .. -- ..

j By letter dated May 13, 1996, the Director, Office of Nuclear Reactor Regulation (NRR), NRC, acknowledged the NRC's receipt of the Petition, and, i for the reasons stated in the letter, denied Petitioner's request for -

immediate action suspending the operating license or reducing the licensed 3

operating capacity of Maine Yankee (Requests 1 and, in part, 4). In addition, i

for reasons stated in the May 13, 1996, letter, the Director denied the Petitioner's request for an informal hearing (Request 5). The Director also j ,tated in the May 13, 1996, letter that the request that the NRC place

'etitioner on service and mailing lists relevant to its interests in safety at '

Maine Yankee and its intention to participate in all public forums opened by the NRC (Request 6) was moot, as Petitioner's attorney had already been added to the Maine Yankee service list. In addition, the Petitioner was informed that NRC would review the Petition in accordance with 10 CFR 2.206 and issue a final decision within a reasonable .me.

i The remaining specific requests for NRC action in the Petition dated January 20, 1996, i.e., Requests 2, 3, and 4 identified above, and the issues

, that Petitioner raised as their bases, are addressed in this decision. For the reasons set forth below, Petitioner's remaining requests for action pursuant to 10 CFR 2.206 are denied.

3 4

l II. DISCUSSION The NRC staff has conducted a thorough evaluation of each of the two safety-related issues raised in the Petition regarding the adequacy of the containment and reactor coolant welds. Each of the issues is addressed below.

1

r

a. Adequacy of Containment Design at or Above Originally Authorized Power Level The Petitioner asserts that the containment is inadequate for operation at any power in excess of that authorized in the original license, and may be inadequate for the originally licensed power level because of insupportable origin;l design acceptance criteria in that the Maine Yankee containment was designed and constructed without diagonal rods. The Petitioner states that "the Atomic Energy Commission staff recommended to the commission that a license amendment permitting this type of construction be allowed, ' ...for this plant and this plant only du, to low seismic risk.' Early in 1979 the MYAPS was shaken by an earthquake of 4.2 magnitude and epicentered less than ten miles from the plant site. The NRC then ordered the shutdown of five nuclear power stations including MYAPS until piping and ,

piping supports could be seismically qualified..." '

The Petitioner also states that there is no public record, however, that j NRC reevaluated what Petitioner asserts is a marginally acceptable containment design at Maine Yankee before it granted license amendments to operate at increased power.

The Maine Yankee containment is a reinforced concrete structure. The 4

original NRC operating license review determined that the seismic and thermal-hydraulic design of Maine Yankee's containment structure is adequate. (The construction permit for Maine Yankee was issued on October 21, 1968, and the operating license was issued on September 15,1972.) With its Petition of January 20, 1996, the Petitioner enclosed an NRC letter of January 22, 1971, in which the staff asked the licensee to submit additional information related to seismic shear stress, given that there are no diagonal seismic shear reinforcements in the containment wall. Low seismicity of the site was not a factor in the staff's acceptance of the Maine Yankee containment design

without diagonal seismic reinforcement bars. As described below, acceptance by the staff of the adequacy of the seismic design was based on the results of stress analyses.

1 The earthquake for which Maine Yankee was originally designed - termed a Safe Shutdown Earthquake (SSE) - is based on a Housner design response spectrum with a zero period peak horizontal ground acceleration of 0.10g. The five plant shutdown that was ordered on March 13, 1979, was triggered by a finding of an error in a piping computer program, which led to the issuance of j IE Bulletin No. 79-07, " Piping Stress Analysis of Safety-Related Piping" on April 14, 1979. The earthquakes that occurred near the plant site starting on i

April 18,1979, at 02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 34 minutes universal time, were not a factor in the five plant shutdown that was ordered on March 13, 1979. As a  ;

consequence of the sequence of earthquakes that occurred near the plant in April 1979 and the occurrence of the January 9, 1982, magnitude 5 3/4 earthquake in New Brunswick, Canada, the licensee undertook a seismic analysis program. This program included analyses and upgrading of certain plant components and a reevaluation of the seismic hazard. Thus, the results from the seismic analyses and upgrading program were instrumental in the staff's conclusion that the existing seismic design for Maine Yankee remained adequate. However, following its review of the seismic hazard reevaluation, the NRC staff determined that the appropriate characterization of the ground motion for any future analysis of the plant is a high-frequency peak ground acceleration of 0.18 g anchoring the response spectrum obtained from NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," using the 50th percentile amplification factors.

J Subsequently, in 1986, the Maine Yankee Plant underwent a seismic margin assessment program. The review-level earthquake used in the seismic margin assessment had a peak ground acceleration of 0.39, which is much greater than the peak ground acceleration of the SSE. Th'e seismic safety margin program included a review of the entire plant including analysis and upgrading of certain plant components, such as Main Contro' Board, Control Room Auxiliary Cabinets, Service Water Piping Support and others. As a result of this reassessment, it was established that, with the upgrades implemented at the plant, the Maine Yankee Plant can be safely shut down during an earthquake with a peak ground acceleration of 0.279 In its report " Seismic Margin Review of the Maine Yankee Atomic Power 4

Station" (NUREG/CR-4826, Vol. 2, dated March 1987), the NRC staff also concluded that the overall seismic margin of the plant, including the

)

containment, was well above the 0.18g value and, therefore, no upgrading of the seismic design was considered necessary. Further, in the staff report "An Approach to the Quantification of Seismic Margins in Nuclear Power Plai.ts" (NUREG/CR-4334, dated August 1985), it is also noted that prestressed and reinforced concrete containment structures have a large seismic margin above the SSE level earthquake.

Additionally, numerous tests and studies conducted since the operating license review of the Maine Yankee Plant, specifically on shear stress in biaxially cracked reinforced concrete without diagonal reinforcement bars, have led to the acceptance of specified allowable shear stress by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),

Section III, Division 2, CC-3421.5, for reinforced-concrete containment structures. An analysis of the Maine Yankee containment structure was

conducted in December 1984 by the licensee and submitted on the Docket as an attachment to letter MN-85-27, dated February 5, 1985. The results of the study indicate that the controlling peak ground acceleration value is 0.399 for the ASME Code allowable tangential shear stress caused by the SSE loading i in combination with design-basis internal pressure and dead loads. This provides additional confidence on the ruggedness of the Maine Yankee containment. ,

Based on the above, with regard to the Petitioner's concern about the adequacy of the Maine Yankee containment structural design for earthquakes (seismic), the staff concludes that the Maine Yankee containment is satisfactory and has adequate margin. The NRC staff has determined that the design of the Maine Yankee containment structure without diagonal reinforcement bars is supported by analysis and poses no undue risk to public '

health and safety. Accordingly, Petitioner's requests for NRC action based on the seismic design of the containment are denied.

b. Microfissuring of Low-Ferrite Stainless Steel Weldments The Petitioner asserts that the Maine Yankee emergency core cooling system (ECCS), reactor coolant piping, and other large piping have not been adequately analyzed for materials degradation to ensure integrity at power operation in excess of the originally licensed power level or under accident conditions. The Petitioner states further that the Atomic Energy Commission's concern with " micro-fissures" in reactor coolant system welds led to the appointment of a task force, and prompted studies and reports in 1971 (before heightened awareness of embrittlement phenomena) that concluded that the microfissures would not propagate or grow under foreseeable conditions. The

Petitioner asserts that large pipe welds next to the reactor vessel have endured 23 years of corrosion, stress, vibration, and radiation and may fail, initiating a loss-ci-ccolant accident, or may be subject to thermal shock failure initiated by use of the ECCS.

l In a safety evaluation dated February 25, 1972, the NRC staff concluded i that the low-ferrite stainless steel weldments in large piping at Maine Yankee j are acceptable because the micro-fissures of the type and density found in the  !

low-ferrite stainless steel weldments of the Maine Yankee facility do not ,

significantly impair the strength and capability of the welds, and that  :

removal of the welos and rewelding could introduce other problems of greater safety significance than those resulting from the presence of microfissures. '

This evaluation was based on information provided by Battelle Columbus Laboratories, Stone and Webster Engineering Corporation, and Dr. Ernest F.

Nippes of Rensselaer Polytechnic Institute. Furthermore, the Maine Yankee  :

reactor vessel meets the requirements of 10 CFR 50.61, " Fracture Toughness l Requirements for Protection Against Pressurized Thermal Shock." In additir,a the large diameter pipe welds attached to, or next to, the reactor vessel d.)

l not receive sufficient radiation to cause embrittlement. Finally, Type 316 stainless steel weld material, in which the microfissures were discovered, is resistant to corrosion in a PWR coolant environment, and the vibratory loads are insufficient to be a concern for large diameter piping.

In a letter to the Petitioner dated May 13, 1996, the staff stated that in order to determine if there is any long-term safety significance of the microfissures, the staff will review the inservice inspection results for the welds identified as being susceptible to microfissures. The staff has now  ;

completed its review of the inservice inspection tests results for welds I

susceptible to microfissures. The staff's review confirmed that no unacceptable indications have been observed during inservice inspection. In addition, pressure tests have not identified any leakage. These tests

, indicate that 23 years of plant operation have not caused the microfissures to grow to a size detectable by inservice inspection or through-wall leakage.

Plug sample testing was performed by Battelle, Columbias Laboratories, on the primary coolant system low-ferrite welds (

Reference:

Battelle's report dated September 17, 1971, which was transmitted by the licensee to the NRC by letter l

J dated September 21,1971). As part of the inservice inspection program in accordance with 10 CFR 50.55a(g), the licensee has been performing and continues to perform ASME Code inspections of large piping welds that may have been susceptible to microfissures at the time of construction. Additional plug sample testing would not yield any pertinent additional information and 1

is not needed.

On the basis of the above analyses, inservice inspection, and pressure

)

test results, microfissures are not considered a long-term safety-significant issue for Maine Yankee. Accordingly, the Petitioner's remaining requests for

NRC action based on asserted microfissures in large piping welds is denied.

I l

III. CONCLUSION

As explained above, and as requested by the Petitioner, the staff examined i

the adequacy of containment design and susceptibility of welds to

^

microfissures. For the reasons stated above, no basis exists for taking any further action in response to the Petition. Accordingly, no action pursuant to 10 CFR 2.206 is being taken in this matter.

_9_

A copy of this Director's Decision will be filed with the Seccetary of the Comission for Comission review in accordance with 10 CFR 2.206(c) of the Comission's regulations. As provided by this regulation, this Director's ,

Decision will constitute the final action of the Comission 25 days after

' l issuance, unless the Comission, on its own motion, institutes a review of the l Decision within that time.

FOR THE NUCLEAR REGULATORY COMMISSION

^

YG.c

Frank f 'Mira 1 , Acting Director )

Office of Nuclear Reactor Regulation Dated at Rockville, Maryland this 20th day of November 1996 J

4 4

1 i

i 4

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

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i 41 $ .

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y,% . . $

EDO Principal Correspondence Control ~

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. f Y

y s. .

!FROM:

23 .:

DUE: 04/ W/96 EDO CONTROL GT96181' l DOC DTt 01/20/96 FINAL REPLYr '-

Anno D. Burt Friends of the Coast - Opposing Nuclear Pollution h

! TO:

) William Russell,NRR

(

'FOR SIGNATURE OF : ** GRN, ** CRC NO: ,

a

!DESC: ROUTING:

1 'A j

l 2.206 PETITION TO SUSPEND OPERATING LICENSE OF Taylor

! MAINE YANKEE ' '

I Milhoan'  ;

i Thompson

{ Blaha j Russell, NRR Jj j

Lieberman, OE

.DATE: 03/26/96 TTMartin, RI ASSIGNED TO: CONTACT:

l DRPE }%9 4

(

Oet- eye.

lSPECIALINSTRUCTIONSORREMARKS:

a j NRR RECEIVED: . , APRIL 4 1996 l NRR ACTION: PE:VAM A i

i NRR ROUTING: SSELL j RAGLIA f TRADANI ZIMMERMAN CRUTCHFIELD

B0HRER ACTION

~

DUE TO NRR D:RECTCR'S OFFl' i BY , .

/8. '94

! OGC 001364 4

.