ML18085A863

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Fracture Toughness of Steam Generator & Reactor Coolant Pump Supports,Salem Unit 1. Revision 1 to Technical Evaluation Rept
ML18085A863
Person / Time
Site: Salem PSEG icon.png
Issue date: 11/30/1980
From: Allten A, Dorschu K, Stilwell T
FRANKLIN INSTITUTE
To: Fair J
Office of Nuclear Reactor Regulation
Shared Package
ML18085A854 List:
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-166, TER-C5257-166-R01, TER-C5257-166-R1, NUDOCS 8102230323
Download: ML18085A863 (17)
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TECHNiCAL EVALUATION REPO.RT FRACTURE TOUGHNESS OF STEAM GENERATOR-ANO.*..

. REACTOR COOLANT PUMP SUPPORiS 1

PUBLI c :SERVI CE.ELECTRIC AND GAS. COMPANY.

SALEM.NUCLEAR.POWER STATION UNIT 1-NRC:DOCKETNO~.: 50-27i NR.CTACNO.

07245 NRC CONTRACT NO. NRC-03-79-1.18 Prepared by

~ranktin Research Center

  • . The Parkway at Twentieth Street* *
  • Philadelphia, PA 19103.

Prepared for.

Nuqlear Aegulatory*Commission.

Washington, O.C. 20555 *

,=..

. FRC PROJECT C5257 FRCTASK 166 Authors:.

T.C.Stilwell, A.G.Allten, K.E.Dorsc::~u, P.N.Noell

. FRC Group.Leader: : T. c. sd.1-we11.

~Lead NRC*Engineer:

J. R. Fair

. Revisicni 1, November 1980 This report was prepared. as an account of work sponsored by an..

agency of the United States Government. Neither the United States Government nor any agen~y then~of, or any. of their employees.*

makes any warranty; express~d or*trnplled, or assumes any legal

  • liability or responsibility for ar:iy third party's use, Or the results *Of
  • *. such use,.** of any Information, apparatus, product or. process disclosed In this report, or represents that Its use by such third*
  • party would not Infringe privately owned right~.*

. ~nklin Research C~nter A DiVision of The Franklin Institute 1101! 23'1 3~7 The Ben,.min Fr~n Pe~y. Pt\\il!I.. Pe. 19103 (21~1 &!8*1000 '

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SUMMARY

.INTRODUCTION

'BACKGROUND

  • CONTENTS CRITERIA APPLIED IN THE EVALUATION
  • 4~1. Fracture-Toughness Grouping of Materials Used in Support Construcdon
  • 4.1.1 Criterio~ *
  • 4.1. 2
  • Interpretation *.

4.2 Plant Grouping for Fractur'e-Toughness Ranking of S/G. and RCP.support Structures.*

4.2.1 CriteriOn.

4.2~2 Interpretation.

4.3 Cri_teria for Fracture-Toughness Adequacy of S/G and RCP Supports.

4.3.l NDT*Criteria for Screening.

'4.3.2 Interpret'ation.

4.3.3 Alternative Criteria TECHNICAL EVALUATION.

s.1 Review Procedure and Implementation of NRC Criteria 5.2 *Extent of FRC Review.

. *. ~-

s,. 3 Review Findings 5.3.l Use of Group I Materials in Applications Important to Structural Integrity of Supports

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-- - 5. 3~ 2 ThiC.k sect~o.n use of Gr0.~p II Mateti"als in Applic'ations Important. to structural':

. Integrity

  • S.3.3_ Thin *section (Jse bf Group II'.Matedals in

. ~pplications *Important To Structural

  • Integrity
  • 5.3.4 Use of* Mate.rials Classified Group III by NOREG 0577, Upon Condi.tiOn.

5.3.5 Use of Materials Cl~ssified. ~;oup III.by NUREG 0577, Outright 6

.. CONCLUSIONS TABLE**

Number.*

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. COMPONENT S_UPP6RT.

SUMMARY

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SUMMARY

Information concerning aspects of.the fracture'"'.toughness. design of, the steam generator (S/G) *and reactor coolant pUin~ (RCP) supports for the Salem Nuclear* Power Station Unit l *was submitted to The Director of Nuclear Reactor Regulation by the Public Service Electric and Gas Company (PSE&G) by letter*

dated.Dec. 30, 1977.

This information was reviewed at the Franklin.Researc})

Center. (FRC) and eval1Jated iri. accordance* with the.,criteria of the Nuclear.

Regulatory Commission*(NRC) as set forth in NUREG 0577-Draft (henceforth

. referred to simply as NUREG 0577).

  • The *information. had. previously been reviewed,ai; part of the. preparation of NUREG 0577, and Salem uril.t l had been assigned a Group III (relatively best) plant ranking for fracture toughness of S/G *and RCP supporti;.. Th~s ranking
  • was regarded as tentative.. Subsequently*, the NRC requested FRC to. conduct an independent review prior to finalizing the.ranking.

FRC's r~view,wasconfined to fracture"."~oughne~~ issues in supports aboye the embedment.

The review was conduct~~fi~ 'accordance ~ith* NRC criteria and to a procedure standardized for the s.everal licensees whose support designs

  • were reviewed at FRC *.

As a result of its review, FRC confirmed that the Group III plant ranking assigned to Salem Nuclear Power *station Unit l for fracture toughness of S/G and RCP supports is justifiable.

2.

INTRODUCTION

. This. report provides. a technical evaluation of info~mation s*upplied by*

PSE&G with. its* letter of '0ec. 30, 1977, to The Director.of Nuclear Reactor Regulation.: The information.concerns the fracture-toughness design of supports

.for the S/Gs and RCPs for Salem Unit l. The objective of the.evaluation.is to rank the design foi fracture-toughness iritegrity on a relative ~cale in acdor-dance with the grouping scheme and criteria established inNUREG 0577. ~nklin Research *center A OMsion ol 'The Franldin lnslilule

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BACKGROUND During the course of the. ~c licensing review for.two pressurized water

,l reac_to.rs (PWR), North Anna Unfts l and 2,. questions were raised regarding. the fracture-toughness adequacy of certain m~mbers of the S/G and RcP supports.

The. potential *for lamellar te.aring in some suppc)rt member1; was also. questioned~

The* staff's concern in the *North *Alln*Llieensing process was that perhaps not.enough attenti.o~*had been given to the selection of materials.for, and fabrication of, the S/G and RCP supports.

Fracture toughness of a mate;ial is a measure of its capability t6 absorb*.*

energy without failUre or dama'ge~' coeneraily, a material. i.s considered *tough*

when, under stated conditions of stress and temperature, the material can withstand loading to its design limit in the presence of flaws~ Toughness also implies that, under. certain conditions, the mate.rial.has the. capability,***

to 'arrest the growth of a flaw.

A lack of _ade~uate toug.hn.ess (accompanied by.

the combination of low. operating temperature, presence_ of flaws, and nonredun-dancy" of er~ tical support niem~rs) could' i:~sult l.n fai.lur~ of the supp6rt structure under. postulated accident conditions, specifically a loss-of~oolant acCident (LOCA) and safe shutdown earthquake (SSE) * *.

To address fracture-toughness concerns at the North Anna facility, the licensee undertook tests not originally sPecified. and *not included in the relevant AS'IM specifications.

These tests indfcated that material used *.in certain support members had relatively poor fracture toughness at 80°F metal temperature.

In. this case, the licensee agreed to raise (by ancillary. electrical heat) the temperature of the S/G suppo~t beams' i~ question to -a minimum of 225°.F every time, throughout the life of the plant, that the reactor coolant.system, (RCS) is. pressurized above i;ooo psig.

The. NRC staf~*found.. this to be*an acceptable resolution.

Because similar materi.als and designs were used in other plants and be-cause similar problems were therefore possible; this matter was incorporated into the NRC Program for.Resolution of. Generic Issues as "Generic Technical :..

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TER-CS257-166 (Rev. 1)

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Activity A~-12, Potential for Lo~ Fracture TOughness and Lamellar Tearing on*

PWR Steam Geli~rator and Reactor Coolant Pump Supports.*.

Since the original lic;:ensing action (North Anna (Jnits l and 2).*involved only the S/G and RCP -support~ of PWRS, the sta.ff's initial 'efforts were.. di-rected t6ward*e~~~foation of the corresponQing.supportsat other PWR facili-ties.

Bow~yer~ the *staff. has kept i~ mind* the possibility of expanding its.

review.to include other suppOrt structure~ inPWRplants and* ~ilpport_struc tur.es in boiling: water reactor (BWR) plants~.

The. integrity of support. embedments_ was riot questioned during.the.North

,Anna licensing act~on; corisequeritly, emphasi!:l was placed' on resolvi~g the most

  • immediate generic -issue-~whether o*r riot' problems simil~r 'to those uncoyered at NOrthArina exist at* other facilities~* It wa~* the staff's judgment that inclu-sion of an evaluation of suppC;rt embedments in the initial re'view. woµld require
  • de~.ailed, plant-si:)ecific investi,gC!tions that were beyond the scop'e of th~ pre-..

liminary, overall generic review~. Sue~ considerations were deemed.more suited to a subsequent phase when more detailed invest igati.ons of i~dividual plants i

mightbe under.ta~en.

Requ~sts for.. 'information -were sent to* licensees in late 19.77;. responses

.to thes.e req~ests w~r~ received. d.uring 1978.

Sandia.Laboratories in.Albuquerque, New Mexico, was retained to,a5Sist the

  • staff in the* review and ana*lysis of the information received from l'ieensees and applicants.
  • Baseq on analysis of this_ information, the technical studies per-:

formed by Sandia :Laboratories, and review of.the issues by the NRC. staff~ *the*

NRC developed an NRC staff technical p6sitionon these iss\\les,* which is pre_.

sented in NUREG :0577, *Potential for Low Fracture Tough.ness and.Laineilar Tear-

~

ing on PWR Steam Geherato~* and: Re~ct~r too'lant Pump Supports.,*

In addition~* NUREG '0577 establishes criteria for evaluation of the fracture.:..toughness adequacy of. S/G and. RCP supports.

NUREG -0577 also applies certain of these er i teria to the ~ilppo~t. s~ructu~es of 'a number of* PWR plants to achieve plant groupings according to the re,lative fractilre-toughriess i11te-. *

  • grity of these* supports. ~--...

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'l'ER-CS257-166 _(Rev.- l)

The plant r~tings are:

Group I (lowest).

Group II (intermediate).

Group III (highest)

During the generic study, a number of PWR'plants were reviewed for.the fracture-toughness adequacy of their RCP and.SiG designs.

As a result of these. reviews, each p*lant was usigned -a tentative. pl.ant ranking' of either

~roup I, II, or III:.

Several Plants, Saiem Unit* l among them, were tentatively 'rank.ed Group III.* In the appendix to NUREG. 057? prepareq by Sandia Laboratories, who initially established the rankings which ~ubsequently received NRC *staff.

endorseJrient.,.the significance 6f the Group III: ranking is described as:

"considered to be as good* as careful, reasonable engineering practice**

can produce.*

  • However, before fin.alizing the tentati.ve Group' III* ranki_ngs, _the. NR.C
  • requested FRC to conduc.t an independent review of the Group III* plants. ( fn. -

eonjunction with similar FRC. task assi.grUri~'nts to* review the fracture~t,c)ughness.

adequacy of corresponding supports in certain Other plants)' an.d to prepare a Technical Evaluation Report for each plant, presentin9 the.. review findings.

The technical evaluation.reported h_erein applies the criteria of NuREG 0577 to the S/G and RCP supports for Salem Unit 1 to*provide an assessment of

. the fracture-toughness adequacy.of thes.e supports ieadin9 to a plan~ ranking.

4.

CRITERIA APPLIED IN THE EVALUATION

4. l.. FRACTURE-IOUGHNESS GROUPING OF. MATERIALS
  • USED IN ~UPPciRT CONSTRUcTION *.,

4.1.1 Criterion

  • Table 4.6, Material Groups, of.Appendix C to NUREG 0577 groups materials according to their relative. fracture* toughness as:*.

Group I (poorest)

Group II (intermediate)

Group III (best)

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'T'ER-C5257-l66 (Rev. l) 4.1.2 rnterpretation If no *supplementary requirements were called out in the material specifi-cation aimed.at procuring a product with fracture-toughness properties supe-rior to those routinely supplied under the material specification, then the material was grouped in accordance with Table 4.6.

If additional requirements aimed at procuring a product with superior fracture-toughness properties were specified, consider_ation was given to cred-

  • iting this specific material order with an improved ma'terial:-group rating. (

4.2 PLANT GROUPING FOR FRAC'l"'URE-~UGHNESS RA~ING OF S/G AND RCP SUPPORT STRUCTURES 4.2.l criterion Plants are classified on the basis o~ the construction materials used in the supports after giving consideration to the importance of their location and function within the structure, and their consequent importance to support-structure _integrity.

(Refer to pages 5 and 6 of* NUREG 0577, Part I.)

4 ~ 2. 2 rnt*erpretation plants were assigned a plant-group ranking identical to* the material-group ranking of the least fracture-tough material used in the construction, provided

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this usage is important to support integrity.

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4.3 CRITERIA FOR FRAC'J'URE-'!'OUGHNESS ADEQUACY OF S/G AND RCP SUPPOR'Y'S It is the cl~ar intent of NUREG 0577 that licensees demonstrate the

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ffacture-toughne~s adequacy of the S/G and RCP supports or that they take appropriate corrective ~easures to assure their fracture-toughness integrity.

NOREG 0577 provides guidance for such demonstrations.

4.3.1 NDT Criteria for screening NDT + l. 3; +

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  • o NOT is the mean nil ductility tra~sition temperature appr~

prbte to the.material as given by '!'able 4.4 of Appendix C to NUREG 0577. :

o* a is the standarddeviation fo~ the data used to determine NDT as listed in ".!'able 4.4.

o

  • Tsuppo~ts i*s the. lo~est metal temperature* that the support member will ever: experience througqout the plant life w}1en the plant is in an operational state.

In the absence of measured, plant~SPeCifie data, TsupPorts is ta.ken as.. 75°F *..

o The temperature term, 30°F or 60°F, is an allowance for. sec-tion size (30°F f.or thin sectioris *and *60°F. for thick sec;.;*..

  • tions) *

~-~*2 rnterpret~tion

+/-f evidence 1s furnished by the licensee *proving that* other. '.'lalue.s of NDT,

. a r or 'J'

. are actually valid for the. S/G or RCP suppc;r:ts and materi-.*

supports

. a lS in the licensee's plant, such data may be used.

If acceptable alte~nati ve

. evidence. is riot available, the' above-stipulated value.s' ~tiould be used~

4~3.3 Alternative Criteria NUREG.0577 also recognized that."fracture-toughness inte~rity is a complex matter involving a number of.interreiated factors, most of which are pl,ant specific *. consequently, demonstrati6n of compliance with the screening crite-ria is* but* one means of providing satisfactory assurance of fractu.re-toughness

  • adequacy.

.NtJREG 0577 not only recegnizes that other means of showing compliance with I

the intent of mJREG 0577 are possible, *but also offers ex*tel'.lsive guidance re-lating to several approaches by which such a demonstration may be achieved.

  • Because of the plant-specific character that such demonstrations must take.,

NUREG 0577 does not restrict the licensees to any singie approach but, in.stead~.

encourages each licensee to review the fracture.:..toughness adequacy of his S/G.

. and RCP supports and submit evidence of his findings*.:

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TECHNICAL-EVALUATio"N The*inforrn~tion furnished to the NRC regarding:the fracture toughness of, and the po~ential for larnellar tearing in, S/G and R~P: supports at Salem Unit.. l *.

was re~iewed *at FRC.

  • This friforrnation was supplf~d in r~spo_rtse to the *NRC staff'~* generic letter to PWR licens.ees c::onc.erning these. issues.,

A' copy of the staff's riequ~st~fo~~iriforrnation iettet (in. generic form) may be. _found in NUREG

  • os11, Ap~end,ix. s.

.Only fracture to~ghness issues were addressed inthe FRC.review; ~he i:eview.

procedure is described. below.

5. i REVIEW PROCEDURE AND IMPLEMENTATION OF NRC CRIT_ERIA :

The drawings and i~forrnation submitted" were_ first examined to become familiar with the. str~ctural design, material selection*,.* and construction

.. practices.* *Key i terns from this_ information were condensed to. tabular form and are presented in Table **s.1.

  • In *accordaric~.. with* a.review.procedure sta~dardized for' the license.es.

~

.whose plants were evaluated at FRC;the first step~'as*to compile.a list pf

. materi~ls use.d. iri all members significant to the' structural integrity. of the

. S/G and RCP supports *.

  • The listed materials we.re taken from. thos.e ret>erted in

. the response to Item 1 of the NRC'.s request for information, supplemented by a suz::vey of the support drawings for additional materials which might> bf!* incH-cated there.

To each' of the mater iais so.identified; *two*. er iteria

  • tests were applied:~
1. *Tl1e NOT criteria.for screening (paragraph4.3.l of this report).*
2.

The material group.ranking in accordanee.with the procedures of Section 4.1

  • For plants which used them, materials with an.assigned Group I or Group II fracture-toughness rating were further categorized as thick or thin by.using the.. formula shown on. the following page to determine the* section thickness ab6ve-which brittle (plain strain).behavior may be anticipated under dynamic load.. ~nklin Re~eis~c:h Center.

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AVAILABLE Construction* Hateri*la:

A-lb A-441 AISl 4140,

A.ISi 4640 Bolting Materials:

A-194 GK 2.

A-325 A-490 YaacOIDllX JOO Camv*c 200 Welding Hateriala:

Yes*

E7016, 17, 18, E70-Tl,T2' F7l-EL12 fABlllCATlON

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Submerged Arc

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,PROCEDURE.

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TABLE 5.1.

'COHPONENT SOPPORT,

SUMMARY

We*.tinghou*e HEAT

.. TREATMENT Silicon* Killed

+Nol'milized A-441 AiSI 4140 H.T.

to 11 ka~. Yield.

AISI' 4640 Annealed

+cold draWn to 97 ~ai. min.* Y.P.

POST-WELDING TREATMENT LOADING CONDITIONS DL + TL.* -

normal*:

PLANT:. SALEH

'AE.

'P *.,S.E.&G.

NDE*ON HATER.I AL DL + TL 4 OBE - upaet DL + TL + PR -.e111ergency DL + TL + DIE -

faulted DL + TL + PR + DBE - *fa'ulted FRACTURE TOUGHNESS.

TEST A-36* not. in tension

. CVN* on A-44 l

. (20 ft-lb ti 20°F)

SUPPOIRT'.SUPPLIER' '.

'MAXIMUM Al,LOWABL! DESIGN STRESS MEMBRANE 6

  • BENDING '(NORMAL)

Normal:

AISC Allowablee "Upset:

l~JJxAISC Allow-ab lea*.

Emergency:

0.9 s..

Faul te3:

l.o '.s,.*

, METIIODS U,SED TO'.,

,PREVENT LAMELLAR

  • .TEARING, THROUGH THICKNESS

'.Max. Thru *. *

  • Thickrieaa, St res ii 19.23 kai'

.* NDEAND..

  • INSPECTl()NS..

PERFORMED

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  • 70°;' (Minimum o~ra~init 'ten.pentu~t!

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The critical -thickn.ess is given *b~;.

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.where:

ayo is the dynamic yield strength. of the stee_l

  • Kr6 is the nominal, minimum a_ssur-ed f~a_cture *.

tdughness of the steel in acc~rdancewith values

supp~ied by NUREG 0577.-.

tc is the critical thickness.* +/-n members thicker than tc, brittle_ (i.e., plane strain) behavior may._ be expected.

A. similar categorization for Group !II materi,als *was not deemed necessary for purposes of the rev'ie~, because s~ch ~~terials are sanctioned for thick-sect-ion. u5e by virtu~ of th_eir group rating

  • Structural drawings *i.iere th~n examined.-for:.*.
1. All st~ucturailY significant uses of Group* r

. inaterialS.

2~

1111 structurally significant uses of Group ri materials in thick sections.

3. *.. -* Structuraily significant applications c:>f m~terials knowri to be senSitive-to stress

~oirosion eracking or other special failure mechanisms which might mak~ them*prone to bri~tle behavior.

The circumstances associated with such usage were th~n examined.

~

. consideration was given to factors such as:i - direction of loadings' (always

' compressive or sometimes. tensile); strf!ss levels in the member as, indicated in the licensee's response, the presence of. stress raisers 1n member geometries, redundancy of lo~d pat-hs,- and-the like*** ':Applications. judged -to be of problem-*

atic fracture'.toughness were identified for more detailed evaluation at a future date.*

rn addition, iriform~tion furnished on welding and material spec~~icatio~s

  • was.examined for fracture-'toughness implications by a welding engineer and a

. metallurgist, respectively*

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As a. result c;if th~ review findings and ;n.accord~nce with the criteria Pt:~c~dure desc~ibed in Section. 4.* 2 *of. this report i.a. tentative plant. ran.king for.fracture'.""toughn~~s adequacy of SiG and RCl?.supports was.assi.gned.

_5.2 EXTENT O~ FRC REVI:D4 FRC's evaluations were restricted to assessments of the fncture*toughness of supp~frts for* steam gen~ra_tor's and reactor' coolant pumps.

Assessment of the fracture-toughness adequacy of supportS for, the other components.and of the'.

  • embedment was*not'iricluded in the s.cope of. FRC 1s*work. assignment and was.not investigated *

. _The.upper region of the steam generators is also constrained against lateral displa~el'llerit by additional structure.

Drawings showing this* structure and its materials of construction were not provided in the material--fu~nished foi:' review *.

  • FRC~ s evaluations are therefore based upon.the review of all sup-

- port structures other thari these *.

5 ~ 3. *REVIEW.FINDINGS~. :

. *s. 3.1* Use of Group I.Materials.in Applicatidns Important to Structural Integrity of Supports None found.

5.;3.2 Thick Sectfon Use of Group IIMaterials in Applications Important to Structural Integrity None found.

  • - s. 3. 3. Thin Section. Use of Group II Materials in Applications Important. to Structural* Integrity.

- Occasional use of AS'lM A-36 steel ~as found in the Salem support structures~ but.only in applfoations which clearly pose no*fract"1re"."'toughness

-. problems *. Use in prindpal elements of the structure was not foµnd and, in*.

the only applications indentified, the A-36 steel was not subject to tensile

.loads *.

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  • .TER-:CS257-l66. (Rev.' 1) 5.3~4 tise of Materiais Classified Group III by NuREG 0577, :.upon* Condition Major. 'structu'ral members of* both the S/G and RCP supJ)orts are constructed
  • . of AS'IM,A.;;441,. a high"."~trength low'.""alloy *st.eel. Thfs,steel,* *as ro~tinely furnished* from the mill, is ranked Group I;t by NUREG 0577 *. Here however, the.

steel was Ordered silfoon-kill~d, no~m'alizea;: and subject to supplementary re-.

quiremtrnts for Charpy., V".""Notch *testing.

  • These re~irements were added to as-

.. 'sure a mill ptoduct of enhanced f~acture_ t()ughne~s. When.A..;,.441 is* or.dered to such requirements,' the steel' is deemed to merit a' Gt:oup III ranking.

camavac 200.~ an,18% nickel mar aging steel,' is specified fo~ hinge pi~' use '

in the RCP ~u~port structure.

eamavac* 200 is a* mater.ial knownto be suscepti*.

ble to stress corrosion c~acking. ' Because of' this, 'it _is classified as 'a*

  • Group ~I material by NUREG 0577 when *no restriction is placed upon its use.

In the* hinge pin appHcation, however,.the pin~ are not subjected tb te*nsile loads.

  • and must only*sustain i;hear (and possibly bending)* loads up0n occasion.

Under*

these circi.tmstances the piris~re not considered topreserit a fracture-toughness problem and thus, in this application, the steel may be. considered equivalent

. to a Group III steel.*

Corresponding bing~ pins in the** S/G generator support Structure. are 8 l/2

  • inch diameter *. Here.AISI 4640, annealed *and cold drawn to 97 ksi mi.nimwn yield.*

strength, is specified as a replacement steel for a Vasconiax stee.l originally.*

specified. *in this application the AISI 4640 steel is not in.tension but may become occasionally loaded in shear. (w'ith some "superimposed bendi~g) *. Although*

not clas111ifij!d by NUREG. 0577, AI.SI 4640 steel, can in this application be con-i;idered equivalent: to a Group III steel,.i~*. FRC's j~dgement *.

Although Vas~oma~. 200 is not specifically classified..*in NOREG 0577,. vas-coin~x 300 is *. : Because this grade is also sensitive t,o stress corrosion crack~

ing when. used* in. humid. atmospheres* and. subjected.to significant stress, NOREG

  • 0577 classifies it. "as a Group I material for unrestricted use in S/G and RCP.**

supports.

Vascomax 300 is* used in Salem only*. for a 4 inch diameter bolt. which provides hold-down.capability to the RCP under. jet reactions from certain postulated pip_e ruptures*.

In all other circumstances, this bolt: 'remains '

~nklin Research C~nter *.

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TER-C52S7-l66 (Rev *. l)

Unstte.ssed. *.* Thus, in this 'sE>ecific application,* stress corro.sion cracking does not appear* Hkely to pres.ent a probl.em, and the use.of.Vascoinax 300 for this bolt can be Sanctioned.

5. 3. 5. Use __ of Materials Clus'ifi.ed Group III by NUREG
  • 0577,.* Outr~ght All bolting and weldin9 materials.
6.

CONCLUSION

. The_desig~ and construction of supports for steam generators and reactor coolant pumps '.at_ Salem Unit 1 has *been revi_ewed for fracture*toughness adequacy at the FRC.

Criteria for the suitability of materials and construction practices t'or S/G ~nd RCP sti~ports. were prov_ided by the NRC staff, as pub.lished in NUREG.

0577-* Draft *. *In. the revie~i -ge*~eral criteria of N:UREG 0577 were specifically

__ applied.to. inf.Ormation _fornished by Public SE!rvice::Electric and. Gas Company

- * (PSE&G) concerning the supports :in Salem Unit l.

  • .The review was restricted to sup~rts <at:!ove the embedment) for steam:

generators and reactor cooiantpumps.* cOnclusions relating to them do not necessadly extehd to the suppert design of other components

  • In the case.of. Salem Unit 1, FRC_ concludes that: *
1.

Engineering measures taken in supp0rt design, material selectidn,

  • material_ specification,_ material acceptance testing, fabrication

.. methods, and inspections prov'ide reasonable. evidence. that the

  • steam generator support_ structures po~sess ~dequate fracture.

toughness :to meet NRC criteria. for a. Gr()up. III. rating_~

2.

En.gineering.measures taken in the design and cbnstruction*ofthe reactor coolant pump supports provide *_similar evidence to qualify.

them for a Group III rating also..

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3.

The Group III (relatively highest) plant rating for fracture-toughness adequacy of su~ports assigned to Salem O.nit 1

.in NOREG 0577-Draft.is.justifiable

  • I

JAN 3 O 1981 I

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