ML19246C353

Forwards for Commissioner Action Revised Draft Annual Rept Section Re Unresolved Safety Issues

ML19246C353
Person / Time
Issue date:	12/28/1978
From:	Harold Denton Office of Nuclear Reactor Regulation
To:
References
FOIA-79-415 SECY-78-616A, NUDOCS 7907240393
Download: ML19246C353 (45)

Text

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a SECY-73-616A December 23, 1978 COMMISSIONER ACTION MEMCRANCUM FOR:

The Ccmmissioners

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Executive Director for Operations

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FROM:

Harold R. Denton, Director Office of Nuclear Reactor Regulation

SUBJECT:

?EV!SIONS TO ANNUAL REPORT SECTICNS Ri:GARDING UNRESCLVED SAFETY ISSUES PURPOSE:

To respond to Ccmmission recuests regarding SECY-78-616 delineated in Temcrandum from Samuel J. Chilk to Lee V. Gossick dated December 13, 1978.

Discussion-

"r. Chilk's memorandum of December 13, :373 celineated a numcer of acticns requested by the Commission as a result of its review of SECY 72-616 regarding re-porting the orogress of rasolution of " Unresolved Safety Issues" in the "RC Annual Report.

The staff Fas revised tne ::roposed Annual Report secticns that were included as Encicsura 1 'o SECV-78-616 in accordance with the discussions with '.he Ccenissi-:n.

The revised sections are 3ttached.

Marginal mark-ings have been included to indicate where revisions have been made.

The s ecific &ctions taken in res-conse to the numbered requests in Mr. Chilk's Cecember 13, 1973 memorandum are discussed below.

1.

Tasks 12, A-17 and A-26 uere added to the list of " Unresolved Safety Issues" and are discussed in the revised,nnual Report sections.

2.

Exclanations aiong the lines of nose gieen at the Decemoer 12, 1978 Co=1ission ree m g of

.vhy certain cene-ic tasks were not includea as

" Unresolved Safe y Issues" will be orovided.o tne Congress in a NUREG -ecort to accomcany :ae transmittal of the Annual ka:: ort

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9 The Ccamissioners 3.

The definition of an " Unresolved Safety Issue" has been modified as indicated in the revised Annual Report section.

Note that t.*o alternatives are provided that have been suggestec 5y "ommissinaer Bradford and Commissioner a 'nedy.

Either ai.ernative is acceptable to the staff.

4.

ine ii.troductory section of the Annual Report was modifie ' considering the OPE suggestions.

Additional information regarding Task A-30 was provided to Comissioner Bradford's staff by telephone discussions on December 13 and 14,1978.

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Harold R. Centon, Director Office of Nuclear Reactor Regulation

Enclosures:

Revised draft Annual Report sections Comissioners' coments should be prov'ded directly to the Offico or the Secretary by close of Gusiness '4onday, January 3. 1979.

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Comission Staff Office ccments, if any, should be submitted to the Commissioners NLT January a,1979, with an infor ation copy to the Of# ice of the Secretary.

If tne paper is of such a nature that it recuires additional time for analytical review and coment, the Commissioners and the Secretariat snould be apprised of when coments may be excected.

DISTRIBUTICN Cc=issioners Commission Staff Offices Exec Dir for Cperations ACRS AS&LSP AS& LAP Secretariat 402 319

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C; AFT AW;AL REPCRT SECTION ACTICN CN TECHNICAL PRC2LE"5 NRC actions en technical problens related to nuclear power plant safety can take a number of different forms. They can be (1) specific licensirg acticns to resolve a pr:blem experienced or identified at an 0;erating react:r, (2) long term researen programs (3) standards deveic; ment efforts, (4) part of licensing c:nstructicn permit or c;erating if cense reviews, er (5) ceneric reviews of issues that invslve several nuclear ;;wer plants.

Items of the fir:: type above that are detemired to involve a -ajor reducti:n in the degree of ;rotecti:n of the pu lic health and safety are re;crted to Congress ;uarterly as Abnor-al Occurrences (see Cha;ter 7).

Ciscussions of several 3:ditional items involving licensing actions at :;erating react:rs are discussed tel:w u r.d e r the heading of CTHER ACT:CNS.

NRC research ;rograms are discussed in Cha:ter 11 and the devel ; ment Of regu-latory standarcs is discussed in Chapter 10.

LN:E5CL'.'E0 S AFETV :55CES ^L*N Backgr und

n 1377, tre O" ice f Lciear ^eact:r :egulati:n 1::t ) iaso tutec a ;regr3-to de'iae, :ategori:e and a ace :enert: tecrnical activities :n a sjste a: :,

integratec tasis. M e initta; af':r: 3 r.c e r :,is :re;r3m, m itec 'n an -:en-ti'icatice of 133 ;eneric tasus.

• ese tasas ::ver i.ar'ety of :::C:s, se e related :: sa'etj, some rela ted to envie:r enul at:er: 3ne s:re rel3te:

to ' rmag : e recula::ry :rocess.

Sucsecuer: to Sole-*-ting t e NP: :re: ram, : e ::n:ress in :ste :77 3 3c a t :' U: to fac:ude, 5 cre : ner --tr;s, to 3 e-d t e Ener:y :e r ani:=:icn c

3 -ew iecti;r

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" UNRESOLVED SAFETY ISSUES PLAN *

" Sect :n 210. The C:-r'issi:n shall develco a ;13e :reviding for s:ecifi:a-i tien and analysis Of anresolved safety issues relating to Nclear reac :rs and snali take soc 9 a: tion as nay te necessary t: i ;ie-ent c:rre:tive measures with ees;ect to su;n issues. Sucn plan :nali c subm1ttec t0 the Congress on or before January 1,1979 ard ;r gress re;or s snail :e inc!uded in tre annual report to the Comission thereaf ter."

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n response tu tnis re; crit ng re w i remen t, the MC ;rovided a recor* *o :Pe Congress ',9ES-C410) in January 1973 tescribing the ;ereric issues preges, of the O'fice Of % clear Reactor Regul ation trat had teen imple ented earlier in 1977 The *.;R progran provides for tre iden M fication of :ereric issues, tne assignrent of priorities, t*e deve?: cent of detailed Task A:ticr Plans to resolve the issues, project ns of dollar ard an:cwer casts, :On.

ttruing nign level marage ent Oversig9t of task gro:ress, aad ;utlic disse -

irat n of inferrat109 related to *Pe tasas as they ;r0gress. Tae :r grar desce1 bed in % REG-Cal: 15, "c.ever, of considerably trea:er sc:;e : an tre "rresolved Saf ety lssues Dian" required by Se:tten 210.

As acted abcve, the WR pmgram includes other generic tasks af imccriance to accarolishing tre %:C's -ission such as tasks 'Or the resiluti r Of en-vie 0erental issues, for tre devele; rent of improveents in *he reactor i t censing process, for consideration of 'est conservative design cr1*eria r operating limitations in areas wnere overly conservative requirewnts Taj be urne:essarily restrictive or : stly, #:r the maintenance and develo: cent of the NRC staff's capabilities to pe r# r i indepencent audit :a':ulati:ns, and 'cr t*e 3c*ual :er#0r ance Of inte;eatent audit cale 21attens.

'a M nual ecort sect'On is 'i-1ted to describirg t e ;rocress on tra:

"C r t i o n O f

• N e ',Q p r0 g r 3-1 regired *: be re;0r*nd *3

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.% All of the gereriC i5sJes re:cr*ed to tre CCngress last fear in Tl E3-cal 3 were c nsidered as cardidates 'or " Unresolved Safet/ :ssues. ' A syste atic review of !Pese issues was Untertaxen. At an aid in C nducting *his review, 9e t:ct:s addressed :y these issues ere eval.ated from the stan :oint J : net r relative cintribution t3 public risk. This risk-based caara *eri;a t;:n was utili:ed in ::njunction w: n a substantial tocj of additicr.al inf:r at.:n (e.g., neavy wai;nt was given to issues t*st result:d from events nat nave teen -e:crtea to tne Congrees as Abnor al ucturren es) to :etermire weten is-s sues et tre efiniti:n of an ".'aracl ved Sa fety :ssue. ' in;s review resulte" in tre identificati:n of, event:en. resolvec Safet/ Issues.

Tre review Or -

ce:s and tre rationale ':r decisiens regarding particular iss es ire :escrite:

in a se:arate reccrt, C;E3 0513.

Altacugn tre ter- '.'nresol <ed Sa fetf Issue" nas teen in use # r some ti e, and tre C 9gress used tre term to centify these,ss es a: cut.nicn it wisnea ta :e ne:

infor-ed, it aas een fre ueatly m sancers cod.

f a gereric safe / issue

'.i.e.,

a safety issue relating to cre t an one plaat; 1s un res ol ved.

• nea new can 'AC ;r3nt a 11:ense *c ::erate a s:eci'i: nuc' ear

wer clart 'Or ani:n : at issue is relevant?

• ne ans,er is taa :e'ere **e license is ranted *ne 'JC staf' must deter-1ne t at liceas g and ::een :n o' *ne s;e:? fic :lant :an continue pending a generic rescisti:n of t*e iss,e.

• be :sses fer these determinati:rs 'ncl :e One or cre c'

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e issue coes no: a: sly :: Or has been resolwe ':

• re :lant.nce-

. rsideration, (2) interin easures are :etog equirec at.:erstir;

. ants rencing firai resolut; n

' **e ss e, 13 resolu* :n can rea' Mably b3 expected te'ere the plant under c;nsf de-ation beg ns aperation, i

or (4) the likelihood of 00:urree.ce and/or the c:nse;ue ces Of an accident scenario for nnich the issue arder stacy is an impCrtant ::nsideration, is s;a77, lb w-a[ ',

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. The NRC staff's c0rclusions in this regard are subjected to e s:

1ay of the licensing process in individual cases. Specifically, the N C staff's conclusions on individaal applications are reviened by the Advis:ry Corrittee en eactor Safeguards and are specifically addressed in the public hearing

recess (see previous section in this Chapter describir.g t?e licensing process).

The seventeen gerecta issues listed tel:a,ere fetermined te.rres]lved Sa'ety ssues ' These generic issaes are 300ressed by taenty *ao ;eaeric or t*e Resolution a f Seaer': :ssaes. The tasa tases in *Pe 'M Oregr3m urce-; of tre 3cclicat'a generic tasks are ;resi ted in :3rentaeses fol!:airl a

• • e title nf each iss e

'9rae o f tre taerty-wo ;ereric ta s<s id:res s M

• • ese seventeen 'ssues aise teen : ~oleted. 1ereric '3sv A-6 aas :; :'e'.ed ind ::crented in a re; ort. t RE 3- 04 :3, "vark : Cent 31n ent 3 bort Te - Dro-

r3a Safety E,aluation eacrt, in

ece-cer 1977 3eaer'; '3sk A 25 was tr-pleted arc d c rente in 'i;RE3 0224, ' es: tor lessel Pressa-9 *ars' eat ie er' Protection *:r Pressuri:e Water e3c*crs,' 4n Se::te-ter 13 73; 3r0 r

73sn '4-21 e s co l et ed a r d :c umen te d in egulatory L1:e 1.129, '3ae:3nce

'r Residual West Ce oval,' in "ay 1973.

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Water Ha rer - (A-1) 2.

Asy retric Sle=de n Loads en the 4 sctor Ccolant Syste, - (A-2) 3.

Pressuri:ed Water Reactor Steam 3eaerator Tube irtegr ty - (A-3, A 4, A-5)

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EWR Mark I and wark II Fressure Su;;ression Centainments - (A-6 A-7, A-8, A-39) 5.

Anticipated Transients Without Scram - (A-3) 6.

2=R Noz:le Cracking - (A-10) 7.

Reactor Vessel Ma terials Toughress - ( A-ll)

L rect;re Taugaaess of Steam iererat:r sa: 2eact:r ::o:ent -:

L::cris A-12) 3.

jste= Nters

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13. Envirce ental halification of 5sfe f ;eisted Ele:tr':31 :: r :~- t -

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11. Reactor vessel :ressure Transieat ;* tec a -

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12. Desica.nl -est :e-cvai 7e:uire eats

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13. :cn*rol af -eavy. sacs Nese ::ent Eel

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ta*1 n 31ack out - 'A.;4) 21sassicn of escn of the Urresol vec Cafetj !ssues' ' ' :vs.

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WATE't HAM _uED (GE'.ERIC TI.SK t 1)

'aater hanrer events, tnat is, intense pressure pulses in fluid systems, sucn as correnly ex;erienced w en rapidly closing a water faucet, often Occur a

in nuclear scaer plant fluid syste s.

Since 197*, about one curcred incidents irvolving water ha rer in nuclear ;oner react:rs have been re crted. These mater hamer incidents have involved many ty;es of flJid systems irCIudirg steam generatcr feed-rirgs, feed-ater nrd steam st.oly pioing, residual

• eat rencval systems, emergeacy core cool <ng syste-s, contairment spray syste s, and sarvice aater systems. The incideats have bean attributed *o sucn causes as the radtd c:rdensation o' steam ;cckets, ste4*-driven slogs of water, puro start-vo alth ;artially e-Otj 11res, and raoic valve mot 4095.

"ost of t e da age *as teen relativelj mince, n: wever, : Pere nave :een several :ases of fa11ere or partial f ailure of system ;1oing.

N: =ater hamer incident has resulted in the release of radicactivity Out-side of a plant. HC=ever, the princi;al c ncerns are that water ha.~er could result in the f ailure of a pice in the reactor c:elant system or

able a system required to 0:01 the plant af ter a react:r shutd

.n.

Naes to prevent one particular type of =ater haver caused by the rapid c:ndensation of steam in the steam generator feed-rings of some pressuri:Pd water reactors are t?ing instituted. A;;11 cants with new steam geaerat3r desigas are being required to demcnstrate tnrcugh test or analysis that water ha m er will not occur in these Oesigns. Plants with steam generat:rs of the t:0 feeding type that are subject to water haver, are being re;uired to modify the feed-rings and/or test the systems to assure water hamer will not occur. Other actiers to ccrrect the specific causes of -ater hater identified to-date are also being required.

Ine NRC s*aff's review of this safety issue has been incorpcrated in the NRC Program fer Resolution of Generic Isstes as Seneric Task A-1 as described in a report (NURE3-0410' to Congrest suomitted in January 1978.

• he potential for.ater hamer in various systems is being evaluated and a;;r0;riate re3 () )

y quirements and systematic review procedures are being devel ped to ensure t M f

wa'er hamer is give9 a;;ro;riate consideratica in all areas Of licens1n, re-view. The task also includes a study of potenti,, water hamer phencnena to.

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aid in the development of design and review p.ocedgres. A technical report providing the results of a staff review of ma:ct !.axer events in nuclear pc-er plants is scheduled for publication in Decemtt.- 1978. Issuance of this report completes a mjor subtask of Generic Task A-1.

The remining subtasks are expected to te completed in 1951

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ASYPETRIC CLCWCCWN LCA05 CN TPE DE AC

'R CC.Ali SYSUM (GENERIC TASK A-2)

In tae very onl1*ely event of a ruct.re of the ;r ary ecolant pi:1r; in lign:

.ater reactors, large renuniformly districuted lea:s.0ald te 'mecse a:en *re reactor vessel, reactor vessel internals, and otaer c: ecrents in tre react:r c clant systen. Trese newly identified asy retric 1: ads,.nien result f r m tre ra:id decressurizatien of tre reactor coolant syster, were not c:nsi:ered in tre Original :esig" of scre facilities. The fcrces ass:ciated uitn a Oestalste3 Orean in the rea:ter c clant piping rear !ne *eact:r vessel, for narole, could affect tne integrity of the rsact:r vessel su: orts and react:r :ressore vessel internals. A significant degree of failure of tre reactor vessel sucocrt sys-tem, alcng with im: acting tne internals, nas a :otential for (1) da aging sys-ter.s :esigned to :001 tre core foll0 wing tne ::stulated pi:ing brean, (2) af-fecting the ca;:a3ility of tre : qir:1 reds *o function cr :erij, (3) :araging otner react:r ::clant sjstem cerconents, ard ( A) causing other ru:tures in *ne initially e ronen reactor colant systea picing iceps a-d attac e; syste s.

The NRC staff's review of this safety issae "as been inc:r;; rated in t*e NR:

Pe: gram for Res lution of Genu 1c Issues as Generic Tasx a-2.

This pr:gr a in-cluding the NRC staf f's Tas6 Action Plan f:r 'ask A-2.as des:r1 bed in a report ongress submitted in January 197E.

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This issua.as criginally 'dentified in ay 1975 by the Vir;tnia Ele:tric and w

P *er Cer:any in reistien to its Nor:9 Anna Units 1 and 2 nu: lear p:.er plants.

A servey of all :;eratirg PWR reactors.as :enducted in C:t:ter 1975.nich shc.ed that asycetric blo.dCon 10 ads had rot teen ::rsidered in 9e design of tre rea:-

ter vessel suc;crts for any 0;erating P' R facility. In Jure 1976, tie

',4 staf' requested all 0;erating F R licensees to assess the acequacy cf tre react:r ves-sel sa: ports at their facilities aitn res;ect to these ee.ly-icentified loads.

est licensees witn Westingneuse plants initially pr : sed ar n; rented insere:c w

inspection program (ISI) of the react:r vessel safe-end to erd pi;e = elds in t'eu of providing the detailed analysis requested by the SRC staff. Licensees with Com:usti r. Engf reering plants suonitted a ;robability study in sa;; rt of a conclusi:n that a break at the 1ccation in the piping necessary to Or:da:e the postulated load na: such a Icw precability of Oc:;rrence that no f artne-anal-ysis was necessary.

Licensees with Sab: ck and,iilcox Olr.ts took an a:-

preach similar to ::r: gig " gi rge9 liter g T0: j )r; kh N

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The ' RC staff's review of these pr:;osed alternatives to detailed piant-!;e:1fic analyses has been cer;1eted with the c:rclusion that the alterm tive pr:::sals should not te accepted in lieu :( the re;uested analyses.

all F.R licensees and Acc0rdingly, the MC staf f sent letters on January ?5,1978 t:

a;pli ants stating that an analysis rust te unde-tacen to assess the design ade; acy of the reactor vessel su;;.

and Other structures t afthstand the icads.*en sc:0unt-As part of Task asymetric loss.cf.c olant ecrident f r:es are taken int r:ve araiytical e dels and c: ; uter ::ces de-A-2, the 'JC s taf f.ill review r-vel ;ed by reactor vendors to cs?:ulate asjaretric bl:nd:-n loadings prier te In ad:iti:n, their use by liceasees ard a;p.ican*s in piant-s;ecific araijses.

tre staff will develop esplicit guidelires and acceptance cr teria 'Or t.9e asp-i rretric lead aralyses and wili ::noc+

i;e tres ;retab.li*.y study to c

nfir-the a:e:ua:y of 5*af f Oecisions related to the C ntinued ;e"ati n f ;Ian's for tre interim ;eriCd anile iask A-2, piant-specific -analyses, a*d re:essary pl ant mcci fications are necessa: y.

Plant c ificaticas to assure that tre postulate leads are ac::-redated

• ave teen ir;ie*enteo late ir. t*e construction stage of several pian's 3rd Pave tear for so. e o;erating piants. :Or piarts pre;osed 3rd are unter staff review m

still uncer 0;erating license revie.<, the NRC staff recaires tnat plaat-spec:fic rice saalyses te c:rcieted and any recessary plant mcdificatters fr 'erer*ec to issuarce of an :;eeating license. The generic efferts fer ;resssrfre:.ater reactors unde-Task A.2 are currently scheduled f:r c r:Teti:n in early i:79.

The ?AC staff nas teen inves*igating this ; hen: ena as i a;;!ies to tciling

=ater reactors and ans datvaiaed that asyrretric ios.s are aisa significant and theref:re need t0 :e evaluated 'or *.ese Ic.er press,,re sys*e-5.

T>e staff is currently devel:;irg plans f or ex;5nding Task A.2 to res:ive ents 1ssue for bo ling.ater react:rs.

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. FPESSURIZED WATER REA:MR STEM GENEPATOR TU3E INTEFITY (GENEDIC TA*4 A-3, A 4 a-5)

The heat produced in the res:t:r at a nuclear poner plant is u-d to ::n,ert wa ter into steam.hich will drive the turbire-generators. In pla-ts e ploying pressurized water react 0rs, the prirary ccolant water which extracts heat Ly circulatirg thecugh the rea:*:r c;re is kept under pressure sufficient to pre-

'his high-pressure water passes thrcugh tubes around anich a vent boiling.

secordarj : olant (also water) is circulating, under scre. hat I;wer pressure.

The ater in the sec0ndary system is alicwed to boil and predsce steam t0 drive tre turbire-generatcrs. The assettly in which the trarsfer takes place is t e s*eam generator. The tubes -ithin it are an integral ;ar* cf the pri ary c0olant tou dtry, keeping t.ae radica *ive ;rimary coolant in a closed system and n

is lated 'rtm *he ervir:nrent The ;rir.ary Concern is the ca;atilitj Of steam geaerat:r *abes to maintain their integr4*y :uring nor al o;erati n ard ;cstu-lited accident c:rditi:ns. In additian, the r equiremen*: for increase steam neaerat:r tute ins;ections aad re; airs

• ave resul*ed in si;nificant increases in occa:aticnal esposures to cevers.

A detailed discussion of tae s;ecifi: ;r:ble-s associated with steam gererator tube integrity that were eccsering at o;erating reactors was ;rovided in *re 1977 NP: Annual Ae; ort, page 95-ine infor ation telow is provided to sa;ple-rent ar: ;;date that infor-aticn.

C:rrosi:n resulting in steam ;enerat:r tube wall thinning has teen observed in several Westingnouse and Co-tustion Engineering (CE) plants for a nu-:e-of jenrs. "ajor char ges in treir sec:rdary aater treatrent process essea.tially elimirated this form of degradation. Another major c rrosi:n-related ;re-rc enen has also been Observed in a nurter of plants in recent years, resulting fe:m a build-up of su; port plate corresicn products in the ane;lus betwen *ne

• :es and tha su;;crt plates. This tuild-up eventually ca;ses a diametral re-duction of t tes, called " denting," and defer ation of the *ste tu;; ort plates.

Inis ;renc enon has resulted in otrer asscciated events including stress ::r.

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resion cracking, leaks at the tube /su;;crt plate intersecti:ns aad -tead sec-4, f. g 0 ~'

tion tracting of *ubes which were highly stressM tecause of s;;;crt plate j 13 l

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In w Yantee Atomic Pe=er Plant, both of hicn had operated et:lustvely ith an all

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.s velatile treatmer t (MT) cf tre se::,r.carj cnolant It had teen thougit that this type of treatment enght preclude the centing phenomeren frc, causir; signifi-cant degradation. The significant devel pments in Westirgh use and C:rb.sti:n Engineering steam gererators, since June 1977, were the folicaing:

Continued tute denting at Indian Point Unit 2, San Cnofre Unit 1. Surry Units 1 and 2. Turkey Point Units 3 and 4, and lesser amcunts of denting at a number of other Westinghouse desig ed reactors. 5*eam ;eaerstor replace ent is planned fc-early 1979 or 1930 at Sarry Units 1 and 2.

Peplacement or retubing is also being considered f:r Turney Paint Units 3 and 4 In the interi.m. the units are 0;erating under restrictions im;osed by the N C.

Discovery of s ;;crt piste ;.ra; king (related to :enting} at lrdian Point Unit 2 and San On fre Unit 1.

Pe oval of several tetes and a section of su;;crt plate at :ndian oint Uni; 2 to inves*igate tne potential f:r steam generat r :leanirg rewenied :Ontiraed active care:sion of the su pert plate.

C:ntinuaticn of tube denting at Millst:ne Unit 2 and "aine Yansee and cisc very of ferting in St. Lucie 1.

Millst:re Unit 2, Maire Yankee, and Artansas Nu: lear Ore Unit 2 ha,e remved lucs and ;or:1:ns Of the solid rim in tre a:;ermest su;;;rt plates to redJce the susceptibility of the plates to centing-related cracks (CE cesigns).

Palisaces Nuclear P er Stati:n is sleeving degraded tunas instead of plugging trem. Tnis ;r::ess.est:res the struct ral

• jrity Of *he tubes nile kee;irg

• hem in service (CE design).

Steam generator t 2te degradatm in 3atcock and Wilcox (31W) steam ;e erat:rs nas been..miteu to t"e Oc:ree Nuclear Plant.bere tre first tute leak Occurrec in Joly 1976,.

In the last quarter Of IN6 and tre first quarter in 1977, *ne-e aas a total of seven plant s utd:wns *: piug le ning tutes in tre inree 00:ree units. To-date, la tute leaks, all at t*e C: nee units.

• ave o::urred in N steam gereraters. The maj:rity of tnese leaking tubes ere lec:ted adjacent to the 0;en ins;ecticn lare. Lat ratory examinatien of re-med defective ! bes indicated that the tute failures -eee :aused by the pr:;agation of circumfer-ential f atigue rack:, of unnnan origin, by fl:w-indsced vibration.

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,, The sig9ificant devel:;ments in SW steam geeerat:rs, since % ;i77,.ere *he following:

- Centirued tute leaks at the Cconee units.

- Initiati:n of a demonstratien tute sleeving pro; ram by Duke F wer C; ;any at the Oc: nee units. The tube sleeves niil not serve as part of the primary coolant tcundary tut will be installed to enange the vibraticrai character-istics of the *utes and decrease the dynamic stresses aad the sas:c;tibility of the tubes to fatig.e cracking.

Fellcain; ins;ecti:r, by licensees of their steam ;e erat:rs and the crp'et*:n of any necessary repair progrars, the NRC usually ust a;; rove or c:rcar in the re-start of each of the severely affected f acilities. 70-date, the uni *s se,erely af fected by the t. e deeting have ccm leted irs,:a:tice and e;atr ;r:gra s arc received NRC a;;;r:,31 for ::eraticn f. limited time periods. Safe 0;eraticn is assured by **e i :csition of strict c:nditions on licensed :;erati:n, re-quiring *he pis;gir; af af fected tutes ard res*ri: ting al':.etle ieas ra*es during c;eraticn.

As the NEC staff centinues to closely enit:r, evaluate, and a;;reve 'ne a:-

ce:tacility cf c:ntinued :;eration of plants ex;eriencing stear ;ena-at:r tu:e problems, it has undertaken a nu-Oer of ge-eric revle.s and st. dies as part Of three generic tasks in *be NRC Program for the Oesolutien of Sereri: Esues; specifically, Ceneric Tasks A-3, A 4, ard A-5 en:n directed at : e ; articular

ble?! of.est r.g",0use, C;-tusti

• n Eegineering and 3atc::k and Wilcox plants, respecti<ely.

Order these tasks generic studies will be : rducted to (1) evaluate inservice ins;ecti:n results fr m : eca*ing reacters, (C) evaisate the ::nse:ue'as :(

tube f ailures ;ncer ;os*ulatec a:Cicent c rditiens, (3) evaluate ta:e stru:tural integrity, (4) establish tute plugging :riteria tased n

e..nf r ation, (5) defire the re:uire ents for menitoring setordary c:olant :Pe tstry, (6) evaluate inservice ins:ecti:n *ethods, and (7) review design ir;rc<e eats

re

sed fcr new ?lants. These studies will be used to revise :strent %:C staff require-ents and guidance regarding trese sut3ects. In 3:aition, under Tast A-3, the N:.C staff will review and evaluate the first pre;csed stes generat:r repla:e-i n,

ment :: erat 1:n *o establish h Nt/ criteria and guican:e On a geaeeic basis f:r w

w of subse;uht rgi ektj operatf ons. These ge sric, tasks'are use in the review t

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In *he ce;rse of per' rr.ing large scale testing of an advanced design pressure-su;;ressicn centainment (Mark !!!), a d d ring in-plant testir; of wark I c:n-tainments, new su;;ression ;:o1 hydredynvic loads were identified d.ich had ret explicitly teen included in the criginal Mark I er Maru II :cntairreat design Oas's

'hese additicral lea:s result f rem dyramic e*fects of crywell air and steam being ra;; idly f:rced into the su;;ression ; col (t:rus) daring a ;cstulated LCCA and from sa:;ressicn pcol res;cnse to varicus modes Of safety relief valve

($RV) 0;eration generally associated alth plant transient 0; erst'ng conditicns.

Since inese new nydre jnamic loads had not been explicitly ::nsicered in the

rt;inal design of t e arx ! and Mar ( :: containments, t*e '.:C s ta f f determiae:

w tr.at a :etailed reevaluation of these c:ntainment syste-tesigns.as required.

As a resalt of the need f:r this reevaluation the affected utilities for ed I and Marv II 0.rers' Gr:s;s ind es:". "as ergaged the Seaeeal Electric ad "o: v arr Cr;any as its program ranagee 20th C.aem ' 3rcups develo;e: t.o-; rase pr: grams c nsisting of a saart-term program and a iO*g-term program 'cr esolsti n of tre

31 djramic c

ncerns fcr treir res;ective ::ntaircent designs. The 0.ners' f c: ;re*ersive ex.

Greves' pr: grams consist of amen; et*.er things, a nu-ter O

erimental and analytical programs to establish generic ;cci dyramic leads, icad ccm

1raticns and Oesign criterta.

The NRC staf f as identified and initiated a nur:er of generic tasks to review and evaluate tre results of the "arx : ard varx II 0,,rer's 3r:;; s*.crt-term and larg-term programs to devel ; te *nical ;;;sitions for use in liceasing icticas on individual plants atilizing the " ark : and warx I: :cnt31r ent desi;as. These gener4: tasas are irci ded in the '.:C Pr: gram <er Resolution of Se-eric Issses (des: rite: in NU;E;-0410 as rated atcve). !;:ec i f i c a l l y, iney are Task A,,

vark : 3 hart-Wr Or:; ram; Task a 7, Marx : Lcng-Te m P ;;eam; Tasa A-3, Park !!

Containment Pr: gram, Tasn A-39, CeteWraticn f Safety :elief isi e (5Fil ::cl Cy a a a'i : Leads and Te-;erature Limits f;r SWR Contair.-*nts.

The cojectives of the Mart I Short-Tem Program ere: (1) n enmire the :entair-ment system of eacn 3=R f acility wit'1 a "a r k I :entair-ent geJ.i-a to verify that

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, it would.aintain its integrity rd functional carabil'ty atea setjected to tre most protable hydredjramic 1: ads irda:ed bf a ;;stulated design tisis itss-of-ccolant accident; and (2) to verify that licensed vark I E.R facilities may 00ntinue to cperate safe.y. without und e risk to the health and safety of the public, while a ethodical, c:r;re*ensive L n;-Tern Fre;ran is :nducted. The VC determined that, f:r the Short-Term P cjram,

• aintenar:e o' : ntai-rent interity and f, nctien*.culd be a:e;uatelj assured if a safety fict:r t failure of at least too.ere de.:1strated t0 exist 'or the.ea est struct; a' Or me: nan-s ital c0r;crent in tre "ars c:ntairrent syst?9 (i.e., if the calculate 1 st" esses in all c: ;; rents of tre affected ::.t2inrent structsre aere 5 0.n to te less than cae-half the stress.nich.:ald casse tPe : r;onent to I0se its structural inte:rity)

The ' a.

ncla:ed t*at *Pe Objectives of tre 3hcrt-Tem Pr: gram had teen satisfied and deca ented ?*e basis for this c n:lasion in the I

Cactaia ent 5 Fort-Tera Pe;; ram Sa'ety E val ua ti on "s port, ' V E; '-:3, da w.

ece-ter 1977.(i.e., Task A-6.as co :leted in :ccertee 1977).

The Oo;ectives of tre.*arx, ic n g t e r-1 Fr:gre 1 are-(1) *o estat 'er; design basis leads taat are a;;r:priate for **e articipated life of es:9 ara I34 v

f a:11ity, arj (2) to restare tre ;rt gi*al interded desi;n sa'etj argins fcr asch arx ! L:ng-.'?m Program consists f a series wa rt I c:ntainreet sjste-The w of ajor tasks an1 subtasks.nich are :estgred to ;revide a cetailed tasis for rydredjramic load definiti:n and t*.e ret

• colrc" and ac:e:tance criteria for the stru::.ral assess ent.

The g reric as:ects u.

re varx. L:n;-Term Fr: gram

-t Lni;ue trai si. A;;licati:ns Lide. s: edaled to will be descrited in 3 -

f te n r:lete: in Oct: ee 1973, vd in t*e L:ad efinition e;;rt, s: eeuled :: te c: !eted in Dece-ter of 1973. 5.0secuently, eacn ut litj.itt a ara plant v

.ili ;er'Or* 3 : Tant un10.e analysis asir; 3;; roved i:a: :e'initi:n rc struc.

i L:ag Te-m tural analysis =ceni;ues to cenc strate ::nf:r-ance.itn the Nrv Pr:;t ar stru:tdral a::e:trce : '*eria.

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The schedule: c:r;:letion date f r the "arx wy,.

ciuding the issuarce of license aren: rents and tre irclercntation of any ;Ya6, d q t:i'icaticns necessary to satisfy the Wark I L:ng-Tem Pr:g-ar structural acce;t.

ar e cri te ri a, i s ".,ec e-ter 1030. :n re :gnitien of this 5:recule, a rurte-Of f acilities are a:::: ting their :.n scNedales to it;1e ent antic 1:ated plant nodi-f1Chtions and 91nin1:e tPe p0tential f0r exterded plant Outages Or urs:Pe ;ied 4 lr n

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.p The cbjective of the NPC staf f's ef forts under Generic Task A-3 related to the II Shcrt-Tem Prog *am (STP).as to revie ard evaluate the pool dyratic wart leads astociated with a postulated large less-Of-coolant accident prcposed by the Mark II Darer's Crovo to dete.'?,ine their acceptability for use ir riant unique aralyses. The Marx II Shcrt-Tem Pr: gram as c:rpleted in October 1973 and docu-rented in NUEE3-0437, " wark !! Centair. ent Lead Plant Dregram Load Evaluaticn and Acce;tance Criteria." With re;3rd to tre Mark II L:ng-Term Pr:grar. (LIP), t:,e II ::nfir story eugerirectal ard NRC staf f =ill evaluate the results of the art w

analytical ;regra s to assess *he +argin for selected 1: ads. The Mark II L:n;-

Ter n Pr ;ran is cur-ently schedaled for corpletien in Oct: er ii-3 Under iereric Task A-39, t*e 'J staf f.ill review and evaluate t*e results of a

the Mark I and warx !! C ners' Gr:ap's ex:eri ental and aralytical ;rograms to estat'ish and ;ustify t9e safety relief vala related ;;cl dyr.amic 1: ads for 3d wai t I and " ark I! c:ntain ent designs. The result, cf Ceaeric Tasa A-39

-111 te an inte;ral ; art cf the final acce;tatility of *he Mark I ard

• ark II pressure su;;ressi:n ::ntair ent designs. This ;eaeric task is currently stredJled for c: Oleti0n in "?ce-ter 1979. An ir.terim assessment Cf multiple-c nsecutive SRV discr.arges is :arreatly teia; ;erfared for the c:erating

" ark I facilities to sa:pcrt de'erral of this issue until the ::rpletion of tre Marx ! Leng-Ter-Pr: gram. The review of these assessments is schedaled for c: Oletion in N:,r:er 1973.

ts ti % t/u 6 a 6 r d 402 333

. ANTICIPATC TmS!ENTS WIT-NT SCW (GE'dR:: TA!r A-H Eclear plants have safety and conte:1 syste s to limit t*e corseaserces of te-;;rary Scc-e deviaticns from abnor-al eperating conditiens or " anticipated transients."

noreal o;erating c;nditions ray be min:r; cthers, oc rring less f-eq ently, In so e ant'cipated transteats, r.ay impose significant de ands On plant e;ui? e9t.

rapidly s*utting do n the nuclear reaction (initiating a " scram"), and thus rapidly redscing the generation of rest in tre reacter core, is an 1 portant safety teas re.

If there ere a actentially severe "artici;ated transient" and tre reacter ststd:.n did ot " scram" as desi-ed, tacc an "antici;ated transient-wit *Out-scran,'

syste9 cr AT'nS, aculd have occurred.

This issue has teen discussed throug* cut the nuclear industrj 'cr a nurter of years.

Historically, the regulatory staff nas escluded very ICw ;r:Datility everts fr:m tPe At issse in the ATW5 discussions is nnetner or not the ;r:tatility of design tasts.

an ATa5 event is sof'icteatly low to.arrart tre continuance of tM :urrent staf f Oractice.ith regard to AI 5, i.e., centirse er:!o.vn fr un t9e design tasis fer nuclear ;c.er s' ants because of its 19 ;rc: ability.

Be:ause of tre ;erceived potential fcr seet:as ::nse ence3. esulting from AT4S events, a num:er cf studies have teen undertanen to as'ess t*e ;r tattiities 49:

c0nse;sences of sucn events. Inese stuttes aave been ;t 'crmed by werd:rs, utility ;r:.:s, 3rd by the AEC and NC regulatory staff. The AT'a5 is sue as in-C0r; crated in tre NRC Program fer "es:Iu*ics of Seneric issues (described in UE;-

CA10 as noted at:ve) as Generic Task 8-9.

In Se;teater 1973, the thea E: st f( pt.si d sred WASH-1270, "Tect,ical amrt an h'

Water C:oled N.er Peac*crs," J.iW se' ;

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Aia-fortn staff "ac:e;tance criteria.:

year ;eriod folSing :sbl.;ati:n of tre staf f re;cr:, each of t*e four res:::r

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manufacturees sa:mitted analyses ard su;;;rtirg inf aticq n A'WS.ni c".as re-vie-ed by tae f.P staff and addressed i9 four status re;cets publisred in Cece-:e-t e. ASH-l:70 197L The staf f re;c-ts evaluated tre inf:r-ation for c:n' r acce ::

criteria 3rd notec.aere design cnarges ard additicna aralyses were recuired.

'sticred.rether t'e W staf f's re;; ire erts are ae:es-The vendors and vers aave 4 The industry conteeds that the pretacility of an A7 5 event is sary and justified ic. as to r.ake ATWS eventt significantly less than estiNted by tre NRC sr ff ard 50 r

r, mirar safety concerns in 11gnt ater reacts

erations.

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3ecas e of the c:ntinuing c;ntroversy over the 'aC staff ;csition since its publicati:n in WAEH-1270,.a staff resiew and evaluaticn Of all the infor stiCn available on *he subject of ATWS, and in particular, the material de<el:;ed d

I subsequent to the ;ublication of the staff statas reports referred to a:cve, was undertaken in the latter part of 1977 and early 1373. A re;crt, '..;E;

460, ass published in *;ril 1973 providtrg the results of this reviea a n d e.31.a tion.

.:.as co^claced in NLPEG-0460 that :ansidering tre ex;ected f reo ency of transients, the reliability of current eeactor scram syste s necessary to meet the safety ot;ectives nas not been demo 'strated and may. ell aa se not teen attained-NLD.EG-046] reconnerded that means of mitigating tre ::n-seo.eaces of ATW: eveats ce ;rovided in plant dasi;ns.

The rec:nrer:ati:ns presented in ';;E3-:460 nave been criticized by industry and scre Te-:ars Of t~e staff as rre:essar11y ::rsersati.e ard there' re *:0 costly. The staff is new e<aluating alternative reans of recscing the ;retability cr : rse uences of ATW5 everts, otrer t*an that rec:nreaded in 'iUPE3-C06 IPe uffectiseness, ccst and other factors such as the effect en the licensing ;ro-cess of t*ese alteraatives is bei' c es aluated. Jas?d cr. this ev aluatier., the staf' will rec:nrend to the C: rissi:r t*e alternatives amicr ;e:vice t~e best talance between safety and :ost f:r re 1esigns, plants oncer c;rstr.:ticn aad c;eratir; plants. The staff es;ects to ;rovice its re::mrendatians to t*e Commissior in earlj 73

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B J NCZZLE C UCKG (rdNEUC TM< A 101 C<er the last several years, inspecti:ns at 20 af tne 23 coiling.:ter react:r (EWR) plants licensed for ;eraticn in the U.S. have cisclcsed s:me degree of ascking in the feed.ater noz:les of the reict:r vessel at all tut t.o f acil-ities. The exceptions were a plant.ith less than one ye r of 0;ecaticn and a plant with welded no:zle thr-41 sleeves. The three other ficisit h! have *3t yet accumulated significant 0;eesting time end aave not yet been inspected.

The feed.ater no::les, part of the "presssre.essel." are an inte;r 1 part cf the pri.ary pressure tc.ndary of ne reactor coalant system and the sec nd bar-rier (af ter the feel cladding) to the release of radicact've fission Or:dits.

All of tre re; aired 3'at 'eed.ater ner:les net the ASE press re vessel code limits, no.ever, and no i rediate acticn was necessary. Secause relatively small no.nts of tase metal have been r*,cwed t/ re; ate ope aticns, there *as been ec sigrificant reLct1:n in safe.y argins. Several plaett have re-oved the stainless steel nc::le cladding as a means of eliminatir; :ract initiati:n since the clad thickreis.as rot necessary to meet code rei ' rce-ent require-l ments. Pe serPeless, the c acking is potentially sericus beLuse:

- Excessive crack ;re th could lead to im;4irmer4 of pressare sessel safety margins requiring marc cc plicated repair.ork than si ;1e grirding.

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- The design safety.argin.;21d te -enced by excessive re val Of case mets 1.

- The es;csure to radiaticn of tne ;ersonnel :erf:r-i ; :as:e:tt:n an; ee:ste tasks can he c:nsiderable.

- The re: air of taese kinds of crack s can result

'.m c:nsi teestle s*std:.n time at the plant affected.

I Ine react:r vene:r (the 3erecal Electric Cc ;any) ard the NC ~3ve c:nclated f r a1 ?neir respective stadies that tne cracking is caused by 'isctuaticrs or g

" cycling" cf tne te ;erature on t% inside sarface of the na::les; that the stainless steel cladding exhibited less resistance t: crack initiat1:n

• nan tre anderlying low-al!cy steel; ard that, after initiaticn in 're stainless steel cladding, c acks can be arcoa;sted ty 0;erational startap and snutdc.n cycles or other c;erati:nally-ircuted transie9ts. It? veador has ar O

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6 stainless steel cladding, reduction of the temperature differential at ine c:-

le, or some combination of these. The licensees involved have increased the numter and extent of inspections of feed.ater nc::les, with careful re; air and reinspection nere cracks were four.d.

The vendai advised these licensees to clcsely mcnitcr startap and 55utdcwn procedare: in an effort to substantially redace the time during which cold feef.ater is being injected into the hot pressure <essel.

In a clcsely reiated area, the 'i:C.as inf:r ed in arch 1977 by the Jereral w

Electric C mpany that a crack had teen found in the no::le of the "c;ntrol red drive (CRJ) return line" in a reacter vessel in a foreign country. The CR",

return lire no::les are the c:enings in EWR oressure vessels thr; ugh anich tne hign pressure ater ir escess of that reded to c;erate and 001 the CECs is returne: to the pressure vessel. Later in

• ,a rch, the Philadel;hia Electric C:m;any re;crted t at similar cra:<ing *ad teen f:urd in the CC return line nc::le at its Pea e Sett:m 7,temi: P:wer Station, Unit 3.

The cracks reser ble. those 'O.nd in the 'eed ater ro::les and see ec to te the result 1

of the same kind of cyclic themal stresses t*at -ere causing feehater nczzle cracks. Bot, the 'creign reactor ama t,e Peach Ectt:m Unit 3 rea:t:r are re;resentative of a s all num er of Ba:s nich Oc not nave a t*e.al slee.e in tae C C return line ne: le.

The liceasee re-oved tre cracks in t*e eauh 30tt:m CPC P : 'e by geir: inn :ut t*e ; racked area, the asimum crack dR;tn teia.; 7/5-inct, and returaed t~e anit to 0;eration -ith the !?3 return line " val <ed aut ' and.itt the f b 3-d ;res-sare in tre CC nydraul:c sjste : o ified.

rs;ecticn of*other C C reta n line re::les.*ich ir<c:r;; rated tre-al sleeves r

irdicated that these sleeves may cc te effective in preventing tni: Oracu rg

mencmea

n. Inc Secrgia ?:.ee C:m;ary f:;rd a crack in tae C C retsrn iire no :le at its Hat:n Plant. 'Jnit 1,.nich did have a themal slee<e.

(The p.[e p% C T' N P '1 /G, Ej j s p p {1 s

crack.as recoved, the ".02:le ca;;ed, 1

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. The NRC staff eff:rts relatt

.o tre resolutien of ;tese two sim iar issees regarding nozzle cr. ting in totiir; -ater reactors.ere consolidated into a single staff ef f:rt, Generic Task A-lC, in 1977. Under Ger.eric Task A 10, ite staff issued interim guidance to :;erating plants in a report entitled, '!nterim Technical Re;crt On i!WR Feed.ater and C:ntrol R:d Crise Return 1. ire !.0::le Cracking" in July 1977. The staf f is of ten ret; iring inservice ir.s;ection using if:vid ;enetrant exar.inations at c;erating reactors in ac::rdance with the frecuency, Orc edsres and acceptarte criteria described in ne at:,e re;;rt.

Additieral effo*ts under Generic Task A.1C include f ll:wirg and reviewing ad-varce+ents in (1) tre de.el:peent and testing of effective 'eed.ater r:::le ther al sleeves and s;argers, (2) life-cjtle testing cf :eatain CR: sjster.

valves, (2) tne deveic: en: Of vari;;s feedaater syste?. and 07 syster. Odi-fications, and (4} 're develc; ent :f viable altras:nic system te:rnires by tne nuclear irLstry :: alN reliatie and c:nsisttit early teter.raticn of cracting fr m ?: sit tns exte*1:r *o tre r2act:r vessel.

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Desistance to brittle fracture, a rapidly pr0;ogcting catas*ropnic 'ailure mode, for a compenent contrining flaws is described quantitativelf by a material procerty genera?ly der 0ted as fracture toughness. 'his resistance to fracture, or fracture t:ughres-has dif ferent values 3nd characteristics useading u;on the T.aterial teing c0asidered. Fcr nuclear reactor cressure vessel steels, *nree considerations are imscrtant. First, fracture t0u ? ness incresses with in-cr23 sing teTearature. Second, fractare tougnness decreases with increasing load rates. Third, frac *are *.cughnes, decreases witn ne; tron irradiation.

In recognition of these considerations, pewer reactors 3rc operated it91n rest-tctions ' posed by tre Tecnnical Dec1fications On.re ;ressare during resta; and c oldc.n c;er3 icns. These restricticns assure tPat the eenctor sessel will Nt be subjec*ed t:: that ccettnaticn of ;ressare 3rd temerat;rc

• hat C0uld cause brittle fracture of Pe vessel i f signi'icant flaws in **e vessei matertal exist. Tre effect of eeutr0n radiation On tre fracture tougnress of tre sessel oterial i s acc:.un'.1 'or in develcaing and evising these Tec*-1 cal Scecification limitat'Ons Oser *ne 'i'e of the plant For tre sersice ti es and 0;erating c3rd:*1:as tj;ical Of :arrent 0;e 3*"rg plants, re3C*0F sessel fract;re *JJghress ;r0 erties provide ;e0 gate 'argins o' safety agstrst sessel fsilu e.

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STEAM GENE 0ATOR AND QEACT0a COOLANT o?P SUPP"RTS (GENERIC TASK A-121 Ouring tne course of licensing review for a s ecific Pressuri:ed aater Reactor (PWR) 3 rumter of ;uestions.ere raised as to (1) the acequacy of the 'racture tougnress pre;erties of tre material used to fabric 3te tne re3cter coolant puro and steam gener3 tor su: cr*s and (2) the po*ent131 for failure cue to l3mellar tear'ng of these sar,e su:: orts. The safety ::ncern is Inat, 31;noug*

these su;; orts are designec for worst-case 3ccicent conditicrs, ;cor fracture tougrness or lamellar tearing could cause the su:: orts to fail if severely loaded daring sucn accicents. Su ;o-t failure could conceiv30ly im; air tne effectiveness of systers designed to nitig3te the cr aequerces Of the 3cci-cent. An exarple of 3 postulated event seque.. cf potential concern aula te a large ice bre3k in t e reactor coolant system.nicn seve-ely leads *"e su::cris, folle-ed by a sa;;crt f ailure of suf ficient agnituce *. hat 3 3;cr co orent suc9 as 3 s*eam ;eree-!:r is sevecely dis laced resultir; in

'3 : l a re of the erergeacy c:re cooling system ci:4rg anicn is needec to ;rovice. cl1ag nater to *he care.

Two differect steel speci fications ' ASTM A36-7Ca ar ASTM A572 '33, ::verec u

TCst of the 7 ate *4al used far the su; Crts of the O'aR in ques

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TC 300ress the frac tsre t uganess a.es tien (13rellar te3rtng 1s discussed se:3r3telj telow) tests not originally s;ecified 3nd ret in !*e relev3-t ASTM sce:-'i:3-ti0rs aere.ade On

• rose Ee3ts o' s* eel 'Or aniC9 es:ess m3terial a35 3v311-30le.

T*e t0ug""ess of the 436 steel a3s foo.d 'o te 3deOua*e, but t*e *0s *-

r aess of the A572 steel aas relatively ;ccr 3t an :erattrg tem;er3t;re :- 30':

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• "9 c3se of !9e 2'ad in questian, *re 3: liC3nt Agreed *o 3 lice"se COnditice an?:N s'a*ed tra! 9e WCuld " Mise *be te*;e"3*Jre of the IST* Ai72 te3ns 7 9 **e s*fam Oe"e*at0r sar: Orts to 3 -iritrun *e Cer3ture of CIS'F ariar 1 re3:! r 0olant sy!!er ;ressJr1:atiC9 to levels 3 Cve 1300 sig, 3ssur'ry 3:e;u3t2

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s A censultant was engaged to rcassess the fracture tougnress of tre steam gererator and reactor coolant pump support materials for all operating NR plants and these in the later stages af operating license review.

The staff has completed a review of the materials utili cd in tne su::cets of 34 potentially affe:ted GRs. Based on the consultant's pre-liminary evaluation,.e have deterined that t.here are a;;roximately 15-2')

plants wnose su; ports have cuesti:nable tou;nness.

'ae ex;ect tPat tnese plants may be required to utilize inservice inspection er auxiliary neating Of adequate tougnnes'. ;-c;erties cannot te demenstrated.

Upon coroletion of our generic study, e will docarent tre generic

rase cf the fracture ::ugnress program and.ill tegin to im;lement tre resul ts on a plant-s:ecific Oasis. The generic solution will result in changes to tae Stancard Rev'ew Plan tc ine:r; crate the less nt lear ed f:r use in ',ture !icense revie s.

The staff has conclucec inat c:ntinue: : eraticn (and licens:ng) of PWRs is justifiec :ending : r.oietion of this task ard implementation :f the task resuits. Su;;crt f ailure is nct ex:ected to cc ;r except ncer Ne unlikely ecm:irati:n Of:

(1) The c currence of an init'ating esent (e.g., a large pipe tren) amica has teen :eterminec to be of !cw ;r : ability ' acral ::eratir; s+resses :n piair; are very lew)

(2) The existence of non-resncant sna critical su::crt structural mem er(s) witn icw 'racture !:ugrness inany su:: orts ::ntain re.

duncant me-cers).

(3) The existence o f su::cet struct.ral ente-s at Ocerat te cer-atures low encugh that t e fracture ::ug ress cf tne ser:rt material y

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The second potential concern (i.e., lamellar teariny ) maj a!>o be a pr0blem in those su; port st,uctures similar in design to the aforementicned ?'WR.

However, continued o;eraticn of P'aRs during our continuirg generic review Of this concern is acceptable, based n the fact tha t a revie4 to date of a ;ru-imately 400 related technical doc' 2nts revealed only are instance Of krcan f ailure from lamellar tearing. This f ailure occurred in of ten-stresse<: trucx Drakes. In addition, the ' actors consicered atove for tre fracture tougn-ness cencern, such as icw stresses durirg noral 0;eraticn and tne !cw ;r a-0111ty of an initiating event equally a:pij to this concern, The gereric fract.re toughress ;rogram is ex;ected.: te c rcletec in Augus t 1979. The larellar tearing evaluation is a longer ter ef' rt

1. nd is espected to be c rpleted in 1981.

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M un occurs beneath aelds and is : incically found in rolleMteel clate fabrications. The tearing always lies within the carent plate. Of ten outside tne transfomed

' visible) hect-af fected :One (HAZ) and is geaerally parallel to "e aeld fusion boundary. Lamellar tearing occurs t certain critical jotrts usually within large welded structures inv-lving a hign egree of stiffress and restraint. Aestraint may te effred as a -ast-icti:n of the Mye"ent O f the va ri 0ut joint :OF00 rents

• '.Jt would LOPJI' < OC ur as a result of ex;ansion rd contraction of -eld mtal and adjacent eec' ens tur'ag aelding '"La ellar Tearing in nelded Steel F4cedcation', "be aelding rst'Ntel t,

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'% 3YSTEMS INTE3ACTI^NS C4 NCCLE 2 PC'aER DLANTS (GENERIC TASK A 171

n N0ve*ter 1974 the Advisory Committee On :ea *cr Safeg;ards requested tne NRC staff to give attention to the evalua*':n 0* safe f systern from a malti-dis. i;11 nary point Of view to identify potentially uncesirable interacti0ns between ;lant syste s.

The concern arises because the design 39d anal sis Of f

systems is frequently assigned to teams with fun *ional er;ineerin; s;ec.alties s;cn as civil, electrical, mechanical, or nuclear.

T-a aestion

'e rether t*e ee design and analysis

.crs of these functional s;ecialists is integrated so as to identify adverse interactions tet,een and a ong syste s.

These adve-se events mignt occur because designer m1;nt ot, Or example, as-sure tnat red.ndancy and inde;erdeace Of safety s> tems are ;rovided ander all c:nditi0ns o' 0;erati0n.here reduncan j and inde:e idence is reOuired te:ause

• ae 'uncticnal teams may *ot be a0e;uately ::Ordinated. Sim;lf stated: One le't aand af 70 knca Or understar:.a.at the ri;nt hand is d0ing in all cases anere it is ecessary 'Or the nards to te cocrdi a'ed.

The NRC staf f believes tha* tts current rev'e :rocedures and safet/ :r ter's i

r0 Vide reascra0le assurance 19at ar 3

ertatle level O' redundancj arc iace-

endence is ;r vided f0r sys* ems

• hat are required fOr safe *f.

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'n mid-1377 this tasn (Task A-17) was in1tiated to investigate sys*? s inter-CCti n fr0m *ae :Oint of view Of :Onfir-ing that uur creseat ;r0'=: ares 3C e;tably ac:Ount f0r ;ctentially urdesiracle interact 40as ;etneen 5"" s n";

Iystems.

I Fe NRC sta"'s :arrent t eview ;ro:edures assi;n ;ri arj rescorsiti_iity fra '

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3"alyses 3' sjs*e*s.

Tast 3-17 will pr0vice an irde:ee: eat investi;at'On safe */ 'unc*i0ns anc syste*s "eOui*ed *3 Oer'Orq taese ',nt!' Ors 4r Orcer *s assess tne ace;ua:y of current review pr0C0d;res. This i*ves*103 ':n.' te COnts:ted Oy 3a"dia La0Crat0r es ;9 der C0" tract assistaa:e

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. 26 The contract effcet, Phase of the task, teg3n in M3y 1973 $nd is ex;ected t0 be c r;Iated in Septer.ter 1973. The Phase 4 investigatien is struct:; red to identiff where interactions are possible, between and areng syste'*s aneee these interactir.ns have the potential of negating or sericusly degrading tae perfor-ance cf sa'ety functions. The investigation will then proceed to idea.tify where our review ;-ocedures may n0t Nave pr ;erly ac::vnted for these irteractions. Firally, bis d on a Jete--ination of ;ne Ove' ali si:-

nificance to safety, in a fo!!Ow-05 ?".ase I: Of the task, s;ecific c0r.

rective ratsares will te tas.q in areas wnere tre investi;3ticr. shcws 3 reed.

As ncted 3:0we, the 'i?C staff telieves that its review pr ced;res 39: $c-Ce;t3".ce Criteria c,reentlj previde reason 3 Ie 3ssarance that 3n accept 3 le level O' sjstem re:undarcy and !rce:endence is ;r0vided in ;I ht desig"s 3

and this task is ex;ected to 00ndirm this belie #-

Nonet"eless, teCluse sherse sf st9*s in!E"30ti0rs are ;0tcrtially Of large 3,;nific3rce t: ;lant safet/ this issue 93s been ircladed as an 'Un'es:lved $3fety :ssae.

' no signi#iCant systs inte"30ti0ns are identi#ied in tPe Phase investigaticn described above, as is ex ected, this 1ssee will nct be treated in 5,tse.

06ent re crts 35 an 'Crres Ived $4fety.ssee.'

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ENVIRCWE'JAL CUALIFICAT:CN OF SAFETY RELATED ELFCTRICAL ETJIMENT (GENERIC TASK A-Ea)

Cespite the conservative design, construction and opera'i ig cractices and quality assurance reasures required for nuclear power Jlants, safety sys-tems are installed at nuc! ear plants to mitigate the

.0 % ecuences of p stu-Some postulated accidents cou d create severe envirenrental l

la'ed accidents.

conditions inside of containrent.

T." mst limiting of these accidents are high ene gy pipe breaks in the reactor coolant system piping or in a main steam line. In either of these cases, the release of het pressuri:ed -ater and steam to the centainment creates a hign tem; erat:,,re environment (250 tc ADO *F) at hign humidity (including steam) and pressure (as high as abou. 50 psig).

For scre a;plications, chemicals are added for fission product rercval to the containrent sprays that are used to reduce the pressure in the c:ntainrent.

Additicrally, scre electrical e;uipment is predicted to be sutrerged foliNing a large pi;e break. Thus, the safety equipment is ex;csed to such envircrPe**al c:rditions and ni eis to remain 0;e-able during this :eriod, as well as for the long-term post-a(cident ;eriod.

The NRC require, that electrical equi; rent in safety syste s, pri cipally the emer;ency core cooling system and c:ntainment isolation and clea u:; systers, be envircnmentally qualiiied to assure that th;s equipment aill ;erter-its re-cuired function in the envircrrent associated ai'h such severe accidents 5;e-cif:c electrical equipment of c:ncern during postulated accicent c nditices includes (1) the instrcentation reeded to initiate the safety sy!*.e s and pr: vide diagnostic inf0rrati:n to the plant cperators (e.o., electrical :eae.

trati ns into contain"ent, any electrical connectors to cabling h,ich trars.

mits sigrals, and the instruments themselves), (2) c;" trol pcwer to ot:r c;erators for certain valves (e.g., ECCS and c;ntairrent isolation valves lccated ir side contalnnent), and (3) f an ::aler motors for tecse ;lants that utili:e fin c:alers for c:ntainment 5 eat ra oval.

The current NRC safety review process for auclear power plants includes cri-teria related to the qualification nf certain electrical equi; rent-These criteria re;uire that electrical ecu'r.cnt ir:ortant to safety ust ze wal-ified to f unction in the envircreent that mignt result fr0m var 1cus accite9t ay n%mhp_ru&

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s s c:nditions. Although such criteria have been a; plied to varying degrees since the early days of c:nrercial nuclear power, the details of these criteria have been more clearly defined over the years.

These clarifications of the criteria rave raised 50me questiens 35 to:

(1) the degree to whicn electrical equi; ent used in ol er plant designs (these new operating) is cap 3ble cf withstanding the envirsnrental conditions (pressure, terperature, humidity, steam, cNemicals, vibra-tion, and radiation) cf various accident c:nditions under anich it must function (i the ";ualificaticn of equi;rev.' in these :Ider plants),and (2) the adequacy of test er analy'es c:nducted far electrical ecui; rent in ne.er plants to " qualify" such e;ui; rent as capable of ithstanding tre c:nditions of the envir:nrent er * 'ed by various accidents during whicn the equipment must function (i.e., t.ne "aceauacy" cf q.ali ficaticr tests).

Witn regard to alcer plants, tre 'oll:aing actions have taken ;iace in recent mon tr..

As a result of a Sardia testirg program teing corducted for t*e Of fice of "uclear Regulat:ry Research, a generic safety c:ccern with tre ace;uacy :f envie:n-mental qualificatior Of certain electrical equipment as identi'ied. This issue aas nt;nlignted by a Nove-ter 4,1377 petiticn f r:m the Unica Of C:rcerned Sc'eatists.hicn reques _ed i rediate action regarding 0;eratirg ;cwer reactces :cd,

e licensing acticns for other pre;; sed Olants.

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. Sutsequen' WC staff investigati:ns in res ;nse to this issue have led, as of June '

03, to seven piant shutdowns for corrective action and extended outages for two other plants to make rnodifications. These a:ticns were fcr the rest part a result of a lack cf c0nclusive infor aticn regarding the qualification of certain safety equipment.

l Having identified problems associated wi*h qualification of electrical equip-ment, the NRC conveyed its infor ation to tre licensees of all Operating reactor facilities th cugh an !ns;ection and Enforcement Circultr wnich was issued on May 31, 1972. The pur;cse of this Circular was to ensure that the kncwl +dge gained by the NRC staf f and the lessons learned would be a;;ro-grietely factored into future actions. The NRC staff also has initiated an aupented inspection effort as part of the normal NGC activities. This effort ill concentrate on the ins;e:ti:n of installed safety rala' o e:-

tr' cal equipment ai.d on an audit of the ret;t es for environment qualification.

Additionally, a review of the envirenrental qualification of safety-related electrical equi; tent has been initiated for 11 cperating reactor facilities in tre Systematic Evaluation Pr: gram (SEP). (See Chapter 7 "Ainormal Occur-rences - 1978.")

Witn ragard to the second questien above, the N?C staff has orked ith t*e industry to develoo star dards for e:ui:: ment qualificatien and documentaticn

.hich.ould assure the nign level of equi; rent relia 0ility requi. ed for nuclear 3;;lications. This effcrt has culminated in the develc; cent cf IEEE Eid. 323,

':EEE Standard for Qualifying Class IE E;ui; ment for Nuclear ?cwer Gere*ating Stations." This standard and 'ts ancillary standards have provided the f cal point for the aevel., ment of envir:nrental qua..fication requirements in f 3:ent years.

!EEE Std. 323.

first issued as a trial use standard (IEEE Std..22-j971) in 1971 and later, after satstantial revisten, a* a final standard (IEEE Std. 323-197*)

in 1974 Both version', of the Standaro s~

fcrth basic requirerents for envir:n-nental qualificat:cn electrical equipment but do not ;rev de :pecific cetails for impie entation of these require ents. 5;ecific qualificaticn tecnnicues have pp_ w ~

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hen re4ie.ed and approved by the 9C staff on a case-by-case tasis as a ; art of individual licensing rcticns. These licensing actions include initial c:n-struction permit and operating license a; plication reviews and re;ualificati:n a t'ons for operating reactors where d::gre9tation of the initial ;ualification was not availatle.

The ev;1uticnary r.ature of the ; recess of deveicping ervir nrental ;.alification requirements and the case-by-case ir:le entation of these re;oire ents Fas re-sulted in a disersity Of rathods in use and different levels of doca entaticq of the extent to nich equi; ent is qualified.

Several as;ects of equi; rent ; alification are beiag pursued at thic time by the NRC s'af' and *he nuclear industry On a generic basis to achieve a ncre unif0r-ir;ierentatier Of tre ;s eral quali'icati:n re;uirerents estatiis*ed in IEEI Std. 323 1974 C e such activity is tre dewe'c; ent of interir MC staff ;ositices regar:ir; ":. the requirements of IEEE Standard 323-197t can te ret Th.s activity is a part of Seaeric Tast A-24, "Ervir rrental 0,alifi-cation of Safety Sela:ed Electri:a1 E;uiprent," in the V: Fro; ram for tne Res01a*1:n of Generic !ssues arc is s:*eduled f:r :: ;1etien in i

l Further efforts undar Generic Task A-24 involve the review of the envircrrental qualifica ten ;r:grr 3 of -eac :e veac:rs 3-d arcnitect/ engineers as a t. asis for ;ualifyinc saf ety-relate: electri:al aqui;; ment to the requirements of IEEE.

S tar da rd 223-1974 Perfer-ing tnese reviews an a gene-ic basis rat'er than en case-ty-case licens1r; reviews aill ;rev!:e rescurce savirg; for tre 'F" staf' I

and tre industry. This 'ollow-cn ;crt4-n of the gereric task will be sene:aled folic.inc ::rplet!cn of the devel:; ent c' tha irterim 'JC staff positi 9s re-ferred to above.

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OEACICR VESSO. PRESS"EE TWCENT ;;0TECTICN (CENER!C TASK a-251 Over the past several years, incidents identified as cressure transf er.ts have occurred in essarized water reactces (?WR). To-date, there have been thirty-three such events. Half of these events occurred before the plant achieved in-itial criticality (i.e., before initial operatien of the reactor). The majority occurred daring startup or shutd;.n op trati:ns. All of the pressure transients were such that fracts e mechanics and fatigue calculations indicate that the reactor vessels were not dara;ed Sr.d continued operation of I

these vessels was a :eptable. Nevertheless, the staf f connMed that I

f a;;r:priate regulat ry actions were recessary, (1) to redace the 'requency o

of pressure tr3nsient events and (2) to pr: vide equi;nent which would restrict future trarsients te acceptable pressures. This acticn was necessary because react:r vessel safety rargins would te reda:ed during the lifetire of the vessel dae to neutr:n irradiation causing recxed raterial toughness.

Tte NRC staff's review of this safety issue was incorporated ia the N:C Frcgram for Resolutt:n -f Generic :ssues as Generic Task A-Z5 The final re;crt, %U?EG-02M, "Peacter vessel Pressure Transien' Protec ion far Fressuri:ed Water Reac-

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b k# b Upgradec procetral 0ntrols were i cle ented at 0;erattre Fa'R faci n i Tes wnicn signi*icantly redJced the oc;srrence Of press re traesient events. The fev events nic'1 have cc:yered.ere not sig-ificant and -e-e of the type that will te precluced by eq ii; :ent charges.

• he ajority of the equi; rent c*anges ieptemented at ;erating ?aa 'acilities involve tM addition of a sec:nd 1:wer set point en exist:ng :cwer ::erated relief valves, the additlen cf rew sprin;. loaded relief val <et, or mcdificates to allow ase of existing spring-Icaded reliet valves. A % newly licensed facilities ru.t c c;1ete similar desig-cnanges by their first refuelin; shutt:,n.

The extended equi; ment irclenentation schedule for new f acilities was based u; n the reduced frequer of occurrence of pressure trs%ien' events due to i ;rn ed procedural c:ntrois and the large s;.fety P.argins for new pressare <essels.

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'. RESIDUAL HEAT REMOVAL SPUTDOWN REOUIREMENTS MENERIC TASK A-31)

The safe shutdcwn of a nuclear power plant follcwing an eccident not related to a loss-of-coolant accident (LCCA) has been typically interpreted as achieving a hot-standby c'enditi n (i.e., the reactor is shutd:wn, but system terperatui e o

and pressure are still at or near normal operating values). Consequently, c:n-siderable er;hasis has been placed on the hot-standby condition of a pc*er plant in case of ah ?ccident or abnormal occurrence. A similar degree of e phasis 5as been pt:ced en long-term c:oling,.hich is tjpically achieved by the residual heat removal (RbR) system. The RHR system starts to o;erate. hen the reactar c:clant pressure and te perature are substantially Icwer than their hot-standby c:adition values.

Even though it may gererally be c:nsidered safe to raintain a reactor in a hot-standby 00-dition for a Icng tine, ex;erience shcas that there have been events that required eventual cocid:a and long-term cooling until the reactor cociant system aas cold enough to perf:rm inspection and re; airs.

it is theref:re c::-

vious that the 30ility to transfer heat from the reactcr to the environment i ter a shutd:wn is an irportant sifety function for both :Wo.s and SWRs. Cen-sequently, it is essential that a pcwer plant have the ca;atility to go ir:,m hot-standby to : Id-snutd:.n c:nditions (. hen this is determired to te the safest c grse of acticn) under any accident conditices.

This issue was adopted as a Category A issue and designated as Task A-31, "RhR Shutdown Requirement:' in 1977 It was des:rited in the

• 7C Report to Congress, NUREG-0410, 'NRC Program for the Resciuticn of kneric Issues Related to Naclear TTs

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,rdance with the Task Act I n fo his A, t6 t 1

reauire ents fr/ residual heat re eval systems were translated into ;re:: sed changes to Standard :eview Plan Sec ticn 5.4. 7.

These proposals -ere :ansidere:

by the Regulatory De;uirements Oe.iew C nittee (RRRC) daring its 71st meeting

n January 31, 1973.

The RRRC rec: rended approval af the propcsed changes are further rec:. eaded that (1) the changes be applied on i care-Oy-case b. sis to all c:eratin; reac-tors and all other plants (cust:m or standard) for -nich the issuance of tre c;er-ating license is at:ected before January 1,1979, and (2) the changes te tackh *ted 1k s c)

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-, to all plants (cust:m or staadard) for which construction per.it Or ;rel.:..irary design approval applicaticas mere d::keted bef:re January i,1973, and fcr anicn l

the cperating license issuance is expected af ter January 1,1979. These re :a-rrendations were approved by the Director, NRR and are teing implemented.

d Subsequently, the staff pc itten> :- 4e'y e:uire ents for residual heat rer.0 val systems =ers ircor:0 rated into Regulatory Gaide 1.139,

' Guidance fer Desideal Feat Re eval",.hich was issued fcr public c: vent in May 1978. C:ments ere recei.ed c. ring the latter art of 1975 and it is expected that tnit 'egulatory 3Jide can be issued in its final form in late 1979 or early

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> CCNTRCL OF WEAVV LCACS NEAR SPENT FU Q IG3EQIC TASK A-35)

Overhead hard1tng systecs (cranes) are used to lift heavy cjects in the vicinity of spent fuel in PWRs and WRs. If a neavy objec*, e.g., a s;ent fuel sbic;ing cask or shielding bic:.k. were to fall or tip cnto s;:ent fuel in the storage pool or the reactor core during refueli g and da age the n

fue', there could x a release of radioactivity to the environne'1t and a potential for radiation over-ecesures to inplant personnel. If the dre::ed e

cbject is large, and is assured to drco On 'uel containing a large amount of i

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'ission products with ninimal dec3y time "alculated o'fsi*e doses coul: ex-cees *"e siting guidelire values in 10 CFR Part 100.

Th NRC staff's review of this safety issue has been incer;0ra*ed in the NRC Program for Resolution of Generic :ssues as Gereric Task A-35.

T"e objactive of the tasu is to develop a revisicn to the Star:ard

'e-vien Olan (SRP} based on a reevaluation of cur ent N C requirererts and pr0cedures curretnly utilized at c: erat ng clan *s.

If 'ound to te recess wy, i

tre revisi r will Orovide criteria t: Grther reduce the Dotential *ur eavy leads causing una::e;ta:Te drage to 5:en* 'uel in a st: rage :001 or in "e reactor 00re during re' elirg.

• "e rev' sed 3P,,ill ;rovice t*e basis 'Or i DWentir; addi*i:nal recaire ea*s and :rocedures in ed s*ial :lants wrere aarranted and can te used 19 'uture reytows o' new ;1 ants It is the NCC s'af f's vie.

• hat : rtirued 0: era

• ion cu g our revien Of this ;ereric issue presents nc ;ndue risc *: t*e neal*h and safety 3' the ;3blic.

C;erating f acilities use a variety of design and acministratt 'e easures t:

91ni"ize t"e potential for drc:pirg a heavy CDject over the reactor Ore ;r cver the scent fuel pool. These :esign and a rinistrative measures ha,e teen e#fe0*ive since no aeavy lcad taadling accidents resu1*ing in dama ged 'uel have Occurred in over 3C0 rea:: r jears Of U.S. c erating egerience. Additicral'j,

'Or f acilities that have requested incre'ases in s:ent fuel pcci stcra;e ca;aci*j,

• • e NRC has ?rchibited the movement of. Cats Over #uel asse-l'es in the s;ent fuel :ool that neign more than the ecuivalent eight of cre 'ael assemolv.

Also for

• nose plants where a reviaw c+ can dr 0 or *he crace *and'irg syste-r, year scent 'ael has reen is J' cor:lete, movement of shield i

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l Concurrert with our review, licensees have reviewed their current pr:cedures fer the movemnt of heavy loads over s;ent fuel to assure that the potential n3 handlir.g accident tr.at coulo result in darage of sper.t fuel is minir.iited while our generic evaluati]n proceeds. The majortty of the licensees' submittals of their reviews have been received ard are under review.

Generic Task A-26 is expected M be c:rpleted in early 1979. The 73;k will result in the develecment of generic criteria, nc-ever, imolerect-atica of tnese criteria will be h1gnly de:;endent cn plant design characteristics and the s;ecific crocedures in effect at eacn carticalae plant l

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. SEISw!C ' E5 ?';N CRITERI A (GENERIC TASK 4 40)

NRC Regulations require that nuclear pc.ar plant structures, syrters and : r crents important to safety be designed to withstand the effects of natcral phenocera such as earthquakes. Detailed requiremts and guidance regarding the seismic design of nuclear plants is provided in the NR: regulations and in Regulat:ry Guides is-sued by the Ccmi; tion. He.ever, there are a nuder of plants with c rstruction permits and operating liceases issued tef:re t% 'JC's current replations and rep-latory gui:ance aere in place. For this rear o, rereviews of ehe seismic dest;n of various plants are being undertaken (;rincipally as part of the Comis:f on's Systematic Evaluation Fregram) to essure that these plants do not present an un risk t3 the ;ublic.

The NRC staff is c:nducting Gereric Task A 4C, as ; ort of the UC Prog-am for 5

Resciution of Seneri: :ssues. Task A 00 is, in effect, a cs.r:endium of 5 :rt-te m efforts is s : ort the ree.aluati:n of tre seismic design of operating -eactor, The : jective of Task A 40 is, in ; art, to investigate selected areas of tre sets-design se;uence to detemire t'eir consersatism for al'

~;es cf sites, t:

mit investigate alternate a;pr:acFes to parts Of the design sequence, ar.d to cuantify the overall Conservatism 0* tne tesign se;uence. h this manner this program will aid tne

',R staf f in ;erforMng i*s reviews of *re seismic design of o;erating reac*:rs.

Generic Issk A 40 i! se:arated int ten se:aeste suttasks. The 'aj0ri ty Of the suttasks are screduled f r ::m let n in :e:tercer 12 3.

H.ever, *hree of ; e sattssks relatec : devel :ing state-of *re-art 'ethedaiogy to better tefire eartnquake ;rcur: motion near eartn;uane sour,es are I:n:er te m ef' Orts. ~hese tnree suotasks are scheduled for : r:le*icn in IM1.

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PIPE C?4C(5 T 3CIL!NS ATER PEACTCR$

DERIC TASK A. C)

Pipe cracking has cc:urred in the heat affected renes of welds in primary syste, piping in boiling water reactors since the mid-1960s. These c;acks have occurred mainly in Type 304 stainless steel that is being used in ' cst operating SWRs. The ajor ;reble, is recogni:ed to be intargranular stress corrosion cracking (IGSCC) of austenttic stainless steel c:rpenents that have been made susceptible to this failure re e !j being "sensiti:ed" either by post-weld heat treatrer.: er by sensitizati:n of a narr:w heat affected ::ne I

near welds.

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" Safe ends" (short transition ;f eces tet een vessel n ::las and t*e piping) tnat nr,e been-hi;11y seasiti:ed by furna:e heat treatrent ahile attached to vessels during fabricati:n aere very early (late 1960s) found to be susceptible I

to ISSCC. Se:st.se tney -e-e susceptitie to cracking, the at:T.ic Energy Sm-1i mission ::ck the positicn in 1353, that furnace stasitt:e3 safe ends snruld I

j nct te used on new a;;1icattens..Most of the furnace sensiti:ed safe ercs in cider plants nave been re uvad or clad with a prote:tive raterial, and tnere are cnly a few SWRs tnat still Fave furca:e sersiti:ed safe ends in use. Mest of these, bewever, a-e in smaller diameter lines.

Earlier re:crted cracks (;rior to 1975) :::urred ::rimarily in 4" diameter re-circulation loop-by-pass lines and in IC" diameter core s; ray lines. tre r:cently cracks ere dise <ered in recirculaticn riser pi;ing (12" - 14")

in foreign,;1 ants. Cracking is mest uf ten detected daring :nservice :ns:e -

I tien using ultrasoni: testing technicues. Scre piping cracks have teen dis-covered as a result of prirary c clant leaks.

L ?cause af ne,e,c:arrence, of ear pri ary system cracking, there Pas teer a variety of actions undertaken by t*e NRC.

These act;;,ns in:luded:

1 issuance cf hgulatory Guide 1.44 cn "C:ntrol of the Use of knsitired Stainless Steei" issuance of Re;ulatory Guide 1.45 en " Reactor Coolant Soundry Leak Cetecticn Systems" cloself folic.ing the ine. den:e af : racking in BW0s, including fc mign lr,

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i requirir.g augmented inservice ins;ecticn (additicr.al rcre frequent ulti as:nic examinatico) cf " service sersitive" lires.

i.e., those tFat have experienced cracking a

requiring operading cf leak detection systems Pipe cracking and furrace sensiti:ed safe and cracking has been recently re-pcrted in larger (24" dia eter) lines in a SE-designed.~ in Germany.ith over 10 years of service. iecause the are ends on tnat facility had been fur ace sensitized daring f atricatien, IGSCC.as suspected. As a result of ccncerns regarding these furnace sensitizad safe erds, a safe ead.3s rmed in order to perfor-destructive enemin'ation. During lat ratory esaminat on of the re-i moved safe end, including a s all section cf attached pipe, crack s.ere dis-coverec at,arious Iccaticrs in "e safe ead and in tre eld heat of#ected Icne of the pipe. The cracss in the pipe. eld area were very shallo..itn tne axin.m depth less t an about i m (atout 1/B').

Cracxing in the ';rrace seas ti:ed safe eed was scre nat deeper. The Ge-an experience.as the 'irst kn>n occur-rence of IG5CC ir pi;es as lar;e as 24" in diameter.

In June 197E, a thr:ugn-.all crack.as discy ered in an Inco el recirculaticn riser safe end (10* diareter) at tre Cua e Arrolc 'acility. The crack has tee 9 attributed to IG;C: alt"as;n the aterial in this instance is di"e-ent frcm the Tj;e 2C? stainless steel tFat has been nistorically found to c->ck.

I.b-seneat ultrasonic exrination disc:vered ir.fications in six cf the otrer se ven safe erds. Folicwing tneir re c,al, cracking as disc:vered in si eignt safe ends The cracking a;; eared to Fave cr.;'nated in a tiget cre< ice irt een tre irsice wall of tee safe end and an intereal the-al sleeve. Ocn :revices are krc., to ernance,1G CC.

Differences in raterials gecretry, stress le,els and crevices ar; ear t: ahe the ;roblem at hane Ars.:.ni:ve to a particular tj;e of recirculation r';er safe end (Type 1). As a result Of ;"is event, ultras,nic examination ef the other Ty;e I safe ends in U.S. SWS.s, i.e., at t*e E.rs.1cx 1 snd 2 f acility,.as c:ndocted. No significant indica tions.e re fc.rd in h I.

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Unit 2 and One indication, was icentified at Unit 1.

Althcu;h this indication is relati',ely minor and is not "re;crtable" pt.rsuant to the NRC Regulations, it is continuing to be evaluated. *he ultrasonic indication which was fourd will be reevaluated at another plant shutdyn scheduled for later in 1973.

In addition to discussiens with General Electric (the react:r vend r)

(

regarding recent pipe cracking ex;erieace, Cerecal Electri: was asked to pr: vide an in-depth re;crt en the significaace of re:ent events regarding current ir;s e:ticn, repair, and replaca ent pr:gra s.

The were also asked tc address any new safety cercerns related to the o.urrence of cracking in large i

rain recirculation piping. Based en inferrati:n presented by "e eral Electric to date, and exteasive staff evalcation, it was c:ncluded Pat tre recent cc:grren:es a w c rst'tute a basis for iTediate concern ab ut piant safety, i

nor require any nea i rediate acti:ns by liceasces.

Tre staf f brie'ed tre CCTissi:r cn pipe crackir.g in EW s On A;;st 31, 1973, and on Septem:er la,1973, re-established an W Oipe C*ack Stud / Ir:4.

The Study Occup will s:ecifically address the f0licaing iss as:

the significance of ;*e cracks dis::vered in large dia eter pi;es reiative to the c:rclusiens ard re::,rendaticrs set forth in the refere-ced re;;rt and in its ir;:le eata*i:n d::urert SSE3-0313, j

resolution of concerns raised :Ter the ability ta use ult-as:nic tecnni;.es to detect cracks in austenitic stainless steel, the significance of t*e cracts ':und in lar;e diaceter seasit':ed safe ends,and any reccTendaticas regarcing the :.rreat 'GC :r: gram for dealing wi th this ma tte*,

the potenti21 for stress ::rrcsi:n crackirg in 435, and the significar:e cf 19e safe ead cracking at Cuare Arncld riist4<e t: similar mate ria l and design as;ects at cther facilities.

'he Study Grcup is sc*eduled to :Orplete its evaluati:n and re;;rt in January 1379. In acdit :n to the Studj Srcup ef fort, the E *as.nder ay sever 31 generic tecr.nical revie efforts regardirg flaw detection nich are ai ed at ir;rt -ing piping inspecticn te:hniques and ae utrerents These

cceric ef' orts ar.d ary folicw-en efforts resulting r

m *re Stadj ;-:up's

• sk a 32, evc.uation will :e ;rc:": crated into a rew Categ:ry A generic task, a

"Pi:e Cracks at 3 iltag Water Reacters."

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i O ' ~e l CCNTAP MENT ewe;GD Cf SUMP RELIABILITY MENEDIC TJ5K A 41) 1 Following a postulated loss-of-c:ciant accident, i.a., a break in t.he reactor coolant system piping, water ficwing frer the break culd be :Clie:ted in the emergency s,rp at th210w point in the contairrent This ater would be re:irculated through the react:r system by the e erger.:y c:re c:oling ;;r;s to raintain c:re c: cling. This water would also te circulated throu;n the f

containrent spray system to re ove heat and fissien pr:dur.ts fr:n he c n-I tainrent. Loss of the ability to draw aatar from the e ergeacy sep could disable the emergency core cooling and containment spray systems. The 1

c nse:uerces of the resulting inatility to caol tre react:r c re or t*e cn-

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tainrent at,cs:here c:uld te reiting cf the core and/or breas in; cf t*e c:n-i I

tain.ent.

l Cre ;ostult.ted reans of losing the atility to draw.at?r from the e er;ency sura could be blocka;e by debris. A princi;al source of su:5 debris :uld be the ther al insulati:n en the reacter c:clant system ; icing. In the esent of a pi irg break, t*e subse;uent violent release of the nign pressere.ater in the reac':r coole.nt system ::ald ri; Of f the insulation in the area cf the break. This debris : uld then be s.ept into the sur;, potertially causir; car. age.

Corrently, re;ulatory positicns regar:ing sep casign are ;r.se ted in ;e;utatory Goice 1.:

u ;s for Erergen:y C re C: lirg and Containrect 5;raj Systers,"

.hicn addresses decris (insulation). The llegulatory Guide rec: e ds, in ad-a dition to providing redundant se;arated surps, that two protective s:reens te pr vided. A lo. a:;r ach velocity in the vicinity cf tre surp is required to allow insulation to settle out before reaching tne sur; screening; and it is re:uired that the sura remain fu ttional assuming that One-talf of the screen a

surface area is blocked. The 'GC staff believes that sur; designs in act:rda*ce with this regulat:ry guide ac e;tably resolve this issue. Scretteiess, the SEC staff is ccntinuing to study the benavior Of insulaticn under pipe break c:cditi:ns to gain a better ur.de*st!nding of tv.a it r.ig,t be*3ve.

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i e s A second postulated.neans of losing the ability to draw water from the emergency sump could be abnormal conditiens in the sump or at the purp inlet :uch as air f

entrainrent, vortices, or excessive pressure drops. These corditions could re-sult in pump cavitation, reduced flow and possible damage to the pumps.

Currently, regulatory positions regarding sump testing are c ntained in Regu-latcry Guide 1.79, " pre-Cperational Testing of Emergen Cooling Systems for Pressurized Water Reactors," -hich addresses the

of the recircu-lation function. Both in-plant and scale model tests have been perfor-ed to demonstrate that circulatien through the surp can be reliably acccmplished.

The NRC staf f believes that sum tested in acc:rdance with this Regulatory Guide acceptacly resolve this issue. As supplerental guidance, the staff, througn a centractor, is studying whether furt"er guidance for the design ar.d reviaw of emergency sur;s to assure adequate hydraulic design can be developed.

The NAC staff initially planned to study the issue of containment emergency sump blockage from insulation 4s part of Generic Task C-3, " Insulation Usage Within Containment." In addition, initial clans were to study the vortex for nation issue as part of Generic Task 3-18 " Vortex Suppression Requirements for Con-tainments." However, cont 51nment emergency surp operability is fundanental to the successful operatica of both the emergency core cooling system (reeded to c:ol the core) and the c:ntainrent soray system (needed to assure contain:ent integrity) folicwing a less-of-crolant accident. Fce t'is reascn, these porti:cs of Tasks C-3 and B-13 have been c:mais ed and elevated to Category A as Generic Task A-43 under the more gereral title of "Contairment E.rergency Su p Reliability."

Secause this action has only recently been taken, a Task Action P!an and schedule fer this task hava not yet been developed.

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STATICN SLACKCUT MENERIC TASK A ni Electrical

'wer for safety :yste s at nuclear p=er plants is sucplied by two redundant and independent divisiens. The systems used to remove d* cay heat to cool the reictor core following a reactor shutdown are included areng the safety systecs that must meet these requirements. Each electrical I

divison for safety systers includes an offsite alternating current (a. c.)

power c:nnection, a standby e-ergency diesel generator a. c. power supply, 1

and direct current (d. c.) sources.

The issue of staticn blackout was originally included as Generic Task S-57 in the NRC -ogram for Resolution of Ge*eric !ssues. The task involves a study I

l of whetner or not nuclear pcaer plants snculd be designed to a:cCrr'edate a l

c0mplete Icss of all

2. c. pc.er, i.e., a less of of' site a. c. sources and i

both cnsite emergency diesel gererator sources. Loss of all a. c. for an ex-tended period of tire in pressurized aater reactors acc mpanied by Icss of the auxiliary fee:,,ater pur;s (usually cee of t o recundan' pr;s is a steam tur-i l

bin < driver ouma that is r.ot de;ercent on a, c. pcotr f,e actuaticr or opera-ticq) could result in an inability t: cool the rentor c:re with potentially serious consequences If the auxiliary feedaater ;u ;5 are dGendent cn a. c.

o P0wer to function, then 7 1:ss of al; a. c. pcwer for an exterced ;eriod Ould of itself result in an inability to c:ol the reacter cure. Althou:n this is a low probability event sequence, it c:uld be a significant contribut:r to risk.

Current NRC safety requirements re;uire as a -J9i'.um that diver *e pcwer :ri ses be provided for the redadant auxiliary.?ed ater ;rcs. As r,c3d at:ye, tr s is n:r-ally at:Or lished by utilizing an 3.

c. pc.eeed elet-xtor driven pep and a reduncant steam turbirt a t ven pt.rn.

One design adedacy of plants licensed prior to a: ;tien of the a i r e-e n ts.

An initial survey of c;erating plants has 5:en c ;1eted nich indic.tes that all ocerating pressuri:ed =a'er reacters have ei*her steam turcine driven or diesel driven auxiliary feedwater pum:s (neither of which are de;erdent an 3. c.

pcwer). This assures at least that some ca;acility exists for ac::nncditing an extended loss of all a. c. ;cwer. Further review of Oider plants in this re-gard a;11 be concucted as part of tne NRC's Syste.stic Evaluat on Peram (see i

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Furtner stsey regarding this is:ue will include determining if requirerents beyond diverse pcwer drives for the auxiliary feedaater ;u :s are needed. Suca requirements might include specific time requirenents far which the plant must be ca;able of acc:medating a station blackout This safety issue was previ:usly included in the NRG ?rograti for t's Resolu*

of Generic ssues as Generic Task 3-57, but has recently been elevated to Ca gory A as Generic Task A a4 Because this action has cnl, recently been taken, a Task Acti:n Plan and schedale for this task have not yet been develc;ed.

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