ML20212D765

Expresses Safety Concerns Re Containment Pressure Boundary, Based on 850916 Sser

ML20212D765
Person / Time
Site:	Millstone, 05000000
Issue date:	11/12/1986
From:	Halapatz J AFFILIATION NOT ASSIGNED
To:	Asselstine J NRC COMMISSION (OCM)
Shared Package
ML20209D299	List: ML20209D294 ML20212D750 ML20212D755 ML20212D765
References
NUDOCS 8701050018
Download: ML20212D765 (9)

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SEP 161985 Docket No.

50-423 MEMORANDUM FOR:

Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM:

William V. Johnston, Assistant Ofrector Materials, Chemical & Environmental Technology Division of Engineering

SUBJECT:

GDC 51 COMPLIANCE - MILLSTONE UNIT THREE Plant Name: Millstone Nuclear Power Station - Unit Three Suppliers: Westinghouse; Stone & Webster Licensing St' age: OL Docket No.:

50-423 Responsible Branch & Project Manager:

LB #1, E. Doolittle Reviewers:

J. Halapatz/8. Elliot Description of Task:

SSER Re G0C 51 Compliance l

Review Status:

Complete The Materials Application Section of the Materials Engi,neering Branch, Division of Engineering, has completed its review of Hillstone Nuclear Power Station, Unit 3 compliance with General Design Criteria (GDC) 51.

Based on the information contained in FSAR amendments through number 14, dated July 1985 and additional information supplied by Northeast Utilities (the applicant) in letters dated December 21, 1984, July 11, 1985, and September 5,1985 and telecon of September 10, 1985, we have determined that Millstone 3 has satisfied the requirements of GDC 51.

Our safety evaluation is contained in Attachment 1. is our SALP input.

(1l William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering Attachments:

As stated cc:

See Page 2

Contact:

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l ATTACHMENT 1 NORTHEAST UTILITIES MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423' MATERIALS APPLICATION SECTION MATERIALS ENGINEERING BRANCH 6.2.7 fracture Prevention of Containment Pressure Boundary l

In a previous SER input we indicated that ferritic materials' that are used in the containment pressure boundary will be reviewed to the fracture t ness criteria for Class 2 components identified in the Summer 1977 Addenda of Section III of tne ASME Code.

For Class 2 components, the fracture toughness criteria in the Summer 1977 Addenda of Section III of the ASME Code permits the materials to be either Charry V notch tested at or below the towest Service Temperature, evaluated to the nil-ductility transition temperature requirements of Table NC-2311(a)-1 of the ASME Code, or evaluated using the fracture mechanics methods contained in Appendix G of the ASME Code.

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Ferritic materials that are in the Millstone-3 containment pressure-boundary were procureo *.o earlier f racture toughness criteria than those in the Summer 1977 Addenda of the ASME Code.

Hence, many materials were not Charpy V notch tested at or below the Lowest Service Temperatu:e To demonstrate that these materials meet the review criteria, the applicant used the fracture toughness data presented in NUREG-0577, " Potential for

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Low Fracture Toughness and Lamellar Tearing on,PWR Steam Generator and Reactor Coolant Pump Supports," USNRC, October 1979 and ASME Code Section III, Summer 1977 Adde'nda, Subsection NC.

This data indicates that all materials meet the nil-ductility transition temperature criteria of Table NC-2311(a)-1 except for ferritic materials in the feedwater line.

The ferritic materials in the feedwater line were evaluated using the fracture mechanics methods in Appendix G of the ASME Code.

The licensee used a lower bound reference stress intensity factor (26.78 ksi /Iri.) for determining the allowable material fracture toughness.

According to Appendix G, the reference stress intensity value used in the analysis would i

f be appilcable for ferritic material at 180*F below the materials nil-ductility transition temperature.

Additional fracture toughness data for materials with similar composition and heat treatment as the Millstone 3 feedwater materials is reported in a text by Rolfe and Barsom titled, " Fracture and Fatigue Control in Structures Applications of Fracture Mechanics" (Prentice-Hall,1977).

This data indicates that the reference stress intensity value assumed in the Appendix G fracture mechanics analysis is conservative.

The crack sizes' assumed in the evaluation were greater than that permitted during the preservice examination of the component j

l and allowed for flaw growth in service.

The fracture mechanics analysis i

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-3 Indicates that the ferritic materials in the feedwater li,ne would meet the safety margins recommended in Appendix G of the ASME Code.

Additional fracture mechanics analysis performed by the licensee indicates that the critical crack size for brittle fracture would be greater than twice the depth used in the Appendix G analysis.

Based on our review of the available fracture data and material fabrication histories, the use of correlations between metallurgical characteristics and material fracture toughness, and fracture mechanics analysis performed by the licensee, we conclude that the ferritic components in the Millstone 3 containment pressure boundary meet the fracture toughness requirements that are specified for Class 2 components by the 1977 Addenda of Section 111 of the ASME Code.

Compliance with these Code requirements provides reasonable assurance that the Millstone 3 reactor containment pressure boundary will behaveinpno.brittlemanner,thattheprobabilityofrapidlypropagating fracture will be minimized and that the requirements of GDC 51 are satisfied.

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' MILLSTONE UNIT NO. 3 Tracture Toughness Evaluation /

Microexamination of reedwater Pipe Samples Introduction Title 10 of the Code of Federal Regulations lists general design criteria for viclear power plants (Part 50, Appendix A).

During a review o: ths Millstone Unit No. 3 design, the main feedwater piping w-a identified as potentially nonconforming with respect to crit erion 51 (fracture prevention of t he containment pressure boundary).

Af ter a detailed review of the material and its operating environ-ment, it was determined that testing would be desirable in order to obtain bounding fracture toughness data for the installed piping.

This report provides various fracture toug (g

heats of the main feedwater piping materia,hness data for two limiting l installed at Millstone Unit No. 3.

Discussion In response to concerr.s regarding 10CTR5 0, Appendix lA GDC-51, F.illstone Unit No. 3 containment penetsation feedwader piping materia was examined.

Unit No. 3 feedwater lir.es are 20 inch schedule 100, SA-106, Grade B steel.

Two feedwater pipe sar.ples from different heats were tested (HT.

NO. J5341, HT. No. L2220).

Bridgeport Testing Laboratory performed specimen fabrication and charpy impact testing for fracture toughness data.

The charpy impact results are tabolated on Attachment 1.

l Upon receipt of the f ractured charpy specimens f rom Dridgeport Testing, NUSCO Materials Testing Laboratory photographed all specime..

In the "as-tested" or "as-received" condition at approximately IX magnificat1or. (rigures 1-8).

Tracture surface photographs were then taken of all tested specimens at approxinately 4X magnification (r29ures 9-16).

After fractograpny, the specimen halves were checked for hardness or. the notch surfaces.

The results are shown on Attachments 2 ano 3.

NUSCO Haterials Laboratory then generated lateral expans:or. ( A t t: c hme n t s 4,

5) and shear data (attachments 6,7)

The isr. pact energy results generated by Pridceport Tes ti no Labora tory were plctted versus test temperature ( At tachment s 8,

9).

The lateral i

expansion ( A t t a c h.me r.t s Ir, 11) and shear data (Attachments 12, 13) were alsc plotted.

This was done tc allov an estination of ti. 3 nal du tility temperature far the nateria;s e,

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Microexamination of reedwater

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I Pipe Samples March 30, 1984 Page 2 Representative samples were taken from both heats for microexamina-tion.

The two samples were mounted, polished and etched using a solution of 24 nital ( Hitric Acid 024, 984 methyl alcohol).

Mircoexamination was performed on the mounted samplas at 100X magnification (Attachment 14).

To ensure grain structure uniformity sJ from specimen to specimen, 2 charpy specimens from heat J5341 were sectioned, mounted, polished and etched.

The photomicrographs are shown on Attachment 15.

Conclusions a

The samples removed from heats J5341 and L2220 for microexamina-e tion at 100X are typical of as fabricated (extruded) A106, Grade B steel without a grain refinement heat treatment.

e Heat J5341 appears to have been reduced by cold working about 50% subsequent to any heat treatment.

Heat L2220 appears to have been reduced about 90s subsequent to any heat treatment.

The grain structure remains uniform from specimen to specimen on

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0 the basis of the n.ic,roexamination of the 40 F and 150 F test specimens from heat J5341.

Both structures are typical of A106 Grade B steel, All Charpy impact. testing was pe formed in accordance with ASME e

III, Subsection NC.

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Representative Cross Section from Heat J5341 100X magnification 24 Nital Etch

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j July !!,1985 Docket No. 50-423 B11607 Director of Nuclear Reactor Regulation Mr. B. 3. Youngblood, Chief Licensing Branch No I Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20553

Reference:

B. 3. Youngblood letter to 3. F. Opeka, Request for Additional Inf ormation, dated June 6,1985.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 Response to Materials Engineering Questions Attached is the response of Northeast Nuclear Energy Company (NNECO) to the Materials Engineering Branch, Materials Application Section, request for additional information regarding fr.scture prevention of containment pressure boundary. We trust the response will resolve the Staff's concerns regarding this issue.

If there are any questions, please contact our licensing representative.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY et. al.

BY NORTHEAST NUCLEAR ENERGY COMPANY s

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3. e. opeu4 v

Senior Vice President i

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Then personally appeared before me J. F. Opeka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, an Applicant herein, that information in the name and on behalf of the Applicants herein and that t statements contained in said information are true and correct to the best of his knowledge and belief.

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. Fracture Prevention Containment Pressure Boundary 2.5.2.2 Indicate the lowest service temperature for the following components:

8 inch thick, SA 508 Cl.1 feedwater' system flued heads a.

b.

20 inch schedule 100. SA 106 GR B feedwater piping (by Cameron) c.

18 inch x 20 inch SA 234 WPB feedwater reducers d.

18 inch SA 106 GR.B feedwater piping (by USS - Lorain)

SA 216 GR WCB feedwater isolation valve body and bonnet e.

f.

SA 515 GR 70 feedwater isolation valve disc The lowest service temperature is defined in the ASME Code as "the minimum temperature of the fluid retained by the component or alternatively the calculated minimum metal temperature whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pressure." This temperature is the minimum design temperature and should include the worst possible operating conditions (i.e., loss of heating) and the heat input contribution from the plant lighting and the filtration, recirculation and ventilation 1

system (FRVS) equipment. It is not the 90% confidence minimum value, which was reported in the applicant's submittal.

Response

The lowest service temperature for the feedwater system components listed has been estimated to be 740F, As described in NNECO's submittal of December 21,1984 t estimate is a 90%

confidence lower bound of a normal distribution fitted to the applicable building minimum (30) and mean ambient teinperatures. This estimating technique was suggested to NNECO by Mr. 3. Halapatz of the NRC in lieu of a complex and time consuming heat balance. NNECO supports the use of this technique as a reasonable approach to estim.dng the lowest service temperature for these components.

252.3 Indicate te - nil-d.<tility transition temperature, 'sc,urce of data, the thickness rection ecmperatures (per Summer 197/ Addenda to the ASME Cooc, Figure NC-2311 (a)-1), and the allowable lowest service temperature for the following components 18 inch x 20 inch SA 234 WPB feedwater reducers a.

b.

18 inch SA 106 GRR feedwater pip ng (by USS - Lorain)

SA 216 GR WCB leedwater isolation valve body and bonnet c.

t d.

SA SI S GR 70 feedwater isole cion valve dnc These components sere not dneuwed in the applicant's submittal.

I

Responses None of the feedwater system components were impact tested (which is in accordance with the 1974 edition of the ASME Code). Af ter agreement from Mr. J. Halapatz of NRC, NNECO selected a sample of the limiting component for testing.

The feedwater isolation valve disc was not considered as a potentially limiting location since it is not part of the containment pressure boundary at Millstone Unit No. 3.

This is because the feedwater lines are redundantly isolated, as discussed in FSAR Section 6.2 (Page 6.2-16).

Of the remaining components, the 20 inch piping was identified as the most limiting. Thus only this component was impact tested as reported in Appendix F of NNECO's submittal (W. G. Counsit to B. J. Youngblood, dated December 21, 1984).

This testing is considered bounding for the other feedwater system components.

2.5.2.4 Provide justification for considering the nil-ductility transition temperature (NDTT) of 280F for the 20 inch sch 100 SA 106 GR.B feedwater piping (supplied by Cameron). The applicant indicates that the 20 inch schedule 100, SA 106 GR S feedwater li ping (supplied by i

Cameron) will have a NDTT of 280F.

The Charpy impact data provided by the applicant indicate that the NDTT for this piping will be higher than 280F.

This piping was fabricated without a grain refinement heat treatment. NUREG 0377 Table 4.4 assigns a NDTT of 770F for this component in this condition. If the applicant can not provide adequate justification for an NDTT of 280F, the NOTT for this piping should be considered 770F.

i

Response

As repor ted in Appendix F of (4NECO's submittal (W.

G.

Counsil to B. J. Youngblood, dated December 2f,1984), the 20 inch feedwater piping was impact tested to allow a determination of NDTT. Two bounding heats were tested over a range of tempera tures.

In accordance with NUREG 0577 Revision ; (USNRC, October 1983), the minimum NDTT was estimated to be

280F, l

252.5 If the lowest service temperature is determined to be less than the,

l enaternal allowabic lowest service temperature, provide a replacement l

or augmented inservice program to ensure that no cracks of sufficieN sizes emt m the components.

R esponse:

As reported m our submittal of flerember 21, 1984, the actual inwest ser vere temperatures are greater than the allowable ensnimur'.s.' Thus the requirements of General liesign Criterson il are nct m act ordar.rc with the S tandard lieview I'lan. Sec tion 6.2.7, N e visaari r).

In the raw of the feedw ste-syste e.,.spphrahle MMf Oswie require ncnts eini not enviesde snip.s. t te s ting.

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testing was performed on samples from the limiting component. The testing i

showed that the materials used at Millstone Unit No. 3 are acceptable.

l As part of the preservice inspection, all of the welds in the feedwater lines between containment and the first isolation valve have been ultrasonically inspected and found acceptable per ASME Section XI.

In conclusion, no changes to the inservice inspection program are required.

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Septeinber 3,1983 I

Docket No.30-423 B11697 Direct., of Nuclear Reactor Regulation

\\lr. B. J. Youngblood, Chief 1.icensing Branch No. I Division of 1.icensing U.S. Noelcar Regulatory Corninisuon ILashington, DC 20553

References:

(1)

W. G. Counsil letter to B. 3. Youngblood, Response to SER ltern 23, dated Deccenber 21, 193'4 (2) 11.

1.

Youngblood letter to 1.

F.

Ope ka, R ec,ues t for Additional lof orenation, dated June f.,198 5.

(i)

J. F. Opeka letter to B. 3. Youngb!ood, Response to \\taterials Engineering Branch Questions, dated July 11,1985

Dear \\1*. Youngblood:

\\1:llstone Nuclear Power Stat.on, IJnst No. 3 Response to \\taterials Enginerring Branch SER Confirer.atory ite:n No. 25, General Design Criterion (GOC) SI Iri Ref erence ( ! ), Nor t heast Narlear Energy Coinpany (NNECO) subrnetted a response to SCH con!:rnia tor y itein concerning the acceptabihty of Atillstone Unit No. 3 with respect to GDC 51, " Fracture Prevention of Contain:nent Pressure Boundary." Tne report included in Reference (1) concleaded that the contain nent pressure boundar y at '.iillstone Unit No. 3 is acceptable.

In Referens e (3) NNECo prosided additional inforrnation concern.ng cornpliance with GDC 51 in response to a request uritained in R e f erenet-(2).

Represer :.it. <es f r o r. NNECO roet with the NRC Staff on August 21,198 5 to d s-' uv. P - *es; wises pro eided en Ref erer.cc (1). A t that rnceting, it was agreed that t-He!"rerat- ! report does show conipha ve with GDC 51 for all of the conta.

r.s ur " bound.it y w i t ti the esceptiori of certain feedwater systern w ' e p.

st w.e. a gr eer* t h.. t NNECO would perforer. addit;onal calculations

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'c"*l.ater sy steer, e o.r.potectits to show th.it the requirernents of GDC 51 are

.et.

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T'.us the C lass 2 req...r e e.".ts o f t he l 'f 7 7 A d t"..da to A W I % ines til are niet per pa agraphs NC.2 3 3 i.! ind N(:.2132.2.

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If you have any questions or concerns regarding this subenittal, please contact ye our licensing representative directly.

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NORTHEAST NUCLEAR ENERGY COMPANY I

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BY NORTHEAST NUCLEAR ENERGY COMPANY Their Agent 3.GlOpeka i

Senior Vice President

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. Wd By: W. F. Fee Executive Vice President 1

l STATE OF CONNECTICUT )

) ss. Berlin COUNTY OF HARTFORD )

Then personally appeared before ene W. F. Fee, who being duly sworn, did state that he is Executive Vice President of Northeast Nuclear Energy Cornpany, an Applicant herein, that he is authorized to execute and file the foregoing infoirnation in the naine and on behalf of the Applicants herein and that the staternents contained in said infortnation are true and correct to the best of his knowledge artd belief.

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1 MILLSTOslE IINIT NO. 3 i

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Fit ACTilRE MECilANICS liVAllJATION OF TliE MAIN FEliDWATER SYSTEM m

Limit (5)

Minimwn Maxirnum PreserviceIII Postulated Defcct(2) 2KIM+ KIT (3)

KIR(4)

Cornponen t Material Thickness (in)

Defect Depth (In)

(in)

(K51 $

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% USS C l.1 3.62 0.6 2.2 x 11.2 10.7 26.78

'ipe SA196 Gr B

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Gr WPB I3" Pipe sA106 Gr B 1.01 0.10 0.17 18.6 26.78

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i Valve Rodv S A216 j

and Bonnet Gr WCB 1.75 0.09 0.26 13.4 26.78 0.85 j

Valve Div S A 515 Gr 70 2.12 0.12 0.17 24.0 26.78 0.54 J

Since (2 KIM + KIT)is less than KIR, the acceptance criterion of ASMF III, Appendix G is met.

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I Notes (1)

Maximurn Preservice Defect Depth is taken frorn the applicable ASME XI acceptance standard.

These cornponents have passed their preservice inspection.

(2)

Postulated Defect is ASME Ill Appendn G.

For the flued head it n the standard % thickness by 15 tickness sceni ethptical defect. For the pipes, I

reducer and valve disc at is the inaxionurn allowable inservice defect per ASME XI thus allowing for flaw growth in service. For the valve body and bonnet it is the saine percentage of thickness as the pipes to allow for the possibility of larger defects in cast product forens.

( 3) 2Kgu. Kg7 is the strew intensity for the liensting condition. Norinal, upset, test, einergency and f aulted conditions were ronudered and cold startop with au uliar y feedwater was adt nti fied as lines ting.

For conser <atisen, the enaxiinuin operating pressure of !!30 ps was apphed.

Wate r haintner was not considered since it is unlikely in this section of the f(edwater systern due to the piping layout.

(' 1 Kggt is the minianutn anaterial toughness. This valoc is the low er bound and thus unplies ser vice t ein perat u. t, f ar below the nil doctsht y teinperature.

Although tlus was obtained tro n reactor vessel stects,it does apply un< c it is lower thor) values reported for steels of similar coniposition and t. cat treatenent to those used en the feedwater cornponents. For cuarnple, lloite and Barsoin repor t tugher < aloes for A % structural streel on p. ige 10 of their test "Fra r t ur e and Fatigue Control in Structores Applications of Frarture Mc/ha urs" (Prentare. Hall,1971).

r (5) 1.orut Defert Depth is th" approrni. ate depth at which K app K g ig. Tius is repar ted to illmer at" n.e gim te the.. e eptani e i riteriori.

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Director of Nuclear Reactor Regulation Mr. B. J. Youngblood, Chief Licensing Branch No. I Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20535

References:

(1) 3.

F.

Opeka letter to B.

J.

Youngblood, Additional Information Concerning Certain Feedwater System Components, dated September 5,1985.

Dear Mr. Youngblood:

Millstone Nuclear Power Station, Unit No. 3 Response to Materials Engineering Branch SER Confirmatory item No. 28, General Design Criterion (GDC) 51 in reference (1), Northeast Nuclear Energy Company (NNECO) performed additional calculations for certain feedwater system c6mponents showing that the requirements of GDC-51 were met. In discussing the results (Reference 4) with the Staf f Reviewer, Mr. B. Elliot, NNECO has further cor. firmed that a 4.4" deep by 13.2" long, semi-elliptical, axial aligned crack ire the flued head of the main feedwater system has a stress intensity of 6.5 KSI TTn'due to a pressure of l130 psig. Since this is well below the material's critical value of 26.78 KSI W substantial margin against f ailure exists beyond the postulated 2.2" x 13.2" flaw size.

This illustrates added margin with respect to the requirements of 10CFR 50, Appendix A. Criterion No. 51.

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2 If you have any questions or concerns regarding this submittal, please contact our licensing representative directly.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY et. al.

BY NORTHEAST NUCLEAR ENERGY COMPANY Their Agent j f e /= k -

.1. F. Opeka

/

Senior Vice President if;.;f' f ua By:

W. F. Fee Executive Vice President STATE Of CONNECTICUT )

) ss. Berlin COUNTY OF HARTFOR D

)

Then personally appeared before me W. F. Fee, who being duly sworn, did state that he is Executive Vice President of Northeast Nuclear Energy Company, an Applicant herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Applicants herein and that the statements contained in said information are true and correct to the best of his l

knowledge and belief.

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JUL 18 1983 Docket !!o. 50-423 i

ND10RAi100M FCR: Thomas it. Novak, Assistant Director for Licensing l

Division of Licensino l

FROM:

William V. Johnston, Assistant Ofrector i

t'aterials, Chemical & Environnental Technoloqv Ofvtsinn of Engineering

SUBJECT:

GOC 51 COMPLIA'8CE - ?!ILLST0!!E UNIT TifREE i

Plant Name:

Millstone fluclear Power Station - tinit Three Suppliers:

Westinghouse; Stone & Hehster 1.icensing Stage: OL Docket i!o.: 50-423 P,esponsible Branch I, Project Manaqer:

Reviewr:

J. !!alapatz LR 1. E. Doolittic l

P.equested Completion Date: Open Description of Task: SER Pe GDC 51 Compliance Review Status:

Awaiting Conffriatory Information The !*.aterials Engineering Branch, Division of Enqineering with the applicant and architect-engineer, Stone & W I

on June 29, 1983.

, Md.

boundary and the Ifmiting environnental temperature re naterials under operating, maintenance, testing and postulated accident conditions cf ted by GDC 51, " Fracture Prevention of Contaf rvnent Pressu Boundary. "

i Based on the review, the appitcant will provide infortnation conf ts,nin the Ifmitino postulated design temperatures will not violate the limiti a

temperatures identified by the review.

nn pl Wilitan v. Johnston Assistant Director 8'aterials, Chemicel 1, Environmental Technnlogy Division nf Enryfneerfnc At tachnen t: As stated I

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...o ATTACMENT q.,

MILLSTONE NUCLEAR POWER STATI'ON UNIT THREE DOCKET NO. 50-423 MATERIALSENGINEERINGBRkNCH MATERIALS APPLICATION SECTION Fracture Prevention of Containment Pressure Boundary Our safety evaluation review assessed the ferritic materials in the Millstone Nuclear Power Station Unit 3 containment system that consti-tute the containment pressure boundary to determine if the material fracture toughness is in compliance with the requirements of General Design Criterion 51, " Fracture Prevention of Contairvaent Pressure Bounda ry. "

GDC 51 requires that under operating, maintenance, testing and postulated accident conditions, (1) the ferritic materials of the containment pres-sure boundary behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.

The Millstone 3 containment is a reinforced concrete structure, with a thin steel liner on the inside surface, which serves as a leaktight membrane.

The ferritic materials of the containment prfssure boundary which were considered in our assessment are those which have been appifed in the fabrication of the equipment hatch, persor:nel locks, penetrations and fluid system components including the valves required to isolate the sys tem.

These components are the parts of the containment systers which are not backed by concrete and must sustain loads during, the performance of the containment function under the conditions cited by GDC 51

(

i.

We have detemined that the fracture toughness requirements contained in ASME Code editions and addenda typical of t' hose used in the design of the Millstone 3 containment may not ensure compliance with GDC 51 for all areas of the containment pressure boundary.

We have elected to apply in our licensing reviews of ferritic containment pressure boundary materials, the criteria for Class 2 components identified in the Summer 1977 Addenda of Section III of the ASME Code.

Because the fracture toughness criteria, that have been applied in construction typically differ in Code classiff-cation and Code edition and addenda, we have chosen the criteria in the Summer 1977 Addenda of Section III of the Code to provide a unifonn review, consistent with the safety function of the containment pressure boundary mater'ial s.

Therefore, we reviewed the materials of the t

components of the Millstone 3 containment pressure boundary according to the fracture toughness requirements of the Summer 1977 Addenda of Section 111 for Class 2 components.

Considered in our review were components of the containment system which are load bearing and provide a pressure boundary in the performance of the containment function under operating, maintenance, testing and postulated accident conditions as addressed in GDC 51.

These components are the equipment hatch, personnel airlocks, penetrations and elements of specific containment penetratidg systems.

t Our assessment is based on the metallurgical characterization of these materials and fracture toughness data presented in NUREG-0577, " Potential for Low Fracture Toughness and lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," llSNRC, October 1979 and ASME Code Section !!!, Summer 1977 Addenda, Subsection NC.

e The metallurgical characterization of these materials, with respect to their fracture toughness, was developed from a review of how these materials were fabricated and what thermal history they experienced during f abrication.

The metallurgical characterization of these materials, when correlated with the data presented in NUREG-0577 and the Suarner 1977 Addenda of the ASNE Code Section f'I, provides the technical basis for our evaluation of compliance with the Code requirecents

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3 Based on our review of the available fracture toughness data and materials fabrication histories, and the use of correlations between metallurgical characteristics and material fracture toughness, we conclude, contingent on the receipt of confirmatory information, that the ferritic components % the Millstone 3 containment pressure boundary meet the fracture toughness requirements that are specified for Class 2 components by the 1977 Addenda of Section III of the ASME Code. Compliance with these Code requirements provides reasonable assurance that the Millstone 3 reactor I

i i

containment pressure boundary will behave in a nonbrittle manner, that the probability of rapidly propagating fracture will be minimized and that the requirements of GDC 51 are satisfied.

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e Tcblo 4.4 Computation of -NDT Results Material NDT NDT + 1.3a NDT + 20 Cast Steels A-27, A-216 1"

- 6*F 12*F 10*F

,' 4 (heat treated 1"

35 17 57 18'F condition)

^ -

A-352 69 max. -20 Wrought Steels all " mild" steels

• 27 31 67 all " mild" steels 89 except A-201 40 28 77 96 C-Mn*(as-hot rolled) 22 13 39 (normalized)

-28 18

-5 48 8

HSLA* (as-hot rolled) 25**

12**

41**

(normalized)

-50**

18**

-27**

49**

-14**

low alloy non Q&T A-302 8

A-353 28 45 64 A-387 max. -320 65**

Quenched & Tempered A-508 C12 A-514 max.

40*F A-517 max. -10*F A-533B C11 max. -20*F A-537 C12 max.

20*F A-543 max. -60*F s

max. -60*F

{

• • See discussion in Appendix B* See table 3.2 for ASTM specs included in I

i 4.4.3 l

Fracture Toughness e

Minimum values for fracture toughness of the material group l

l s are indicated in Table 4.5.

These are usually dynamic values or static values obtained at lower temperatures equivalenced via the Barsom temperature shift (see section 4.2).

Data at the reference tempera-ture, 75'F, was not,a,1 ways obtainable.

If data was not obtainable, bTR C HM 6eJ T 6""-l, M U(2.6(v og17 iVe M )1"T%4.gwEES OCToW D i f

SES A't rA WCut'G t c-33 T'ott Ew. P W &ri e w j

I

v ww ft 1" section size (Fig. B.1).

Some utility data (Ref. B-1) indicated thick section NDT's in the -100 to -60*F range with a maximum value (one example) of -20*F.

t A-27 Gr 70-40 and A-216 Gr WCB are both C-Mn-Si type alloys varying only slightly in chemical composition allowables, and pri-i marily in minimum yield strength (40 vs 36 ksi, respectively).

Of i

the two, the A-27 Gr 70-40 allows less carbon (.25% vs.30%) but I

L more manganese (1.2% vs 1.0%).

A-216 Gr WCC is virtually identical to A-27 Gr 70-40 in this respect.

A histogram of NDT values for b

r I

A-27 Gr 70-40 heats mainly in the normalized and tempered condition (five were normalized and four were quenched and l

tempered) plus five heats of A-216 Gr WCB is shown in Fig. B.2 This is taken from.a compilation made by the Steel Founder's Society of America (Ref.

B-2).

The statistics of these data imply that 95% of all heats have NDT's below 20*F.

However, these data are taken from 1" thick test castings, and a section size effect may be expected.

[f A second source of data (Ref.

B-3) for these materials indicated that NDT was 35'F l

with a standard deviation (c) of 17'F for 12 specimens of varying thickness (from 2-1/2" to 5") poured from two heats in the normalized j

l and tempered condition.

This still indicates that 95% have their i

I NDT below 70*F, but not with as much margin as the 1 in. thickness 1.

Finally, these two specifications allow the. possibility of V

ccse.

I.l i

producing heats in the annealed condition, if the mechanical proper-HI'l 1

ties can be met.

This would be expected to further degrade their i

frccture toughness properties since a coarser microstructure would l!

recult.

This implies the only way to meet strength requirements

'll would be by increasing carbon content.

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h!ATERIAL (2) Ltaterial Cend.iion Ictters ufe, to:

NC 2310 h!ATERIAL TO BE I.\\IPACT 0 j '"

and Teeuw TESTED HR. He,i noried

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(a) Pressure retaining material shall be impact tested (4) f taterials made to a fine s'aia nieltia2 Practice-in accordance with the requirements of NC 2330 except that impset testing of materials described in (1) through (9) below is not a requirement of this (4) all thicknesses of material for pipe, tube, Subsection.

(1) material with a nominal section thickness e,f fittings, pumps, and valves with a nominal pipe size 6 j

in. diameter and sms!!cr;

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% in. (16 mm) and less where thicknesses shall be (J) material for pumps, valtes, and fittings with g

q taken as defined in (s) through (e) below:

(a) for pumps, valves, and fittings, use the all pipe connections of % in. (16 mm) nominal wall i

thickness and less; largest nominal pipe wall thickness of the connecting (6) austeni ic sisinless sicels; pipes; (b) for veuels and tanks, use the r=minst (7) nenferrous materists; (8) materi-Is li>ted in T:ble NC 2311(.i) 1 fut

-H thickness of the shell or head, as applicable; which the listed value of Tvor* is lower than thP (c) for no:zles or parts welded to vessels, use Lowest Service Temperature * (LST) by an amotnr q

the. lesser of the vessel shell thickness to which the greater than the value of A from Fig. NC-2311(a)-Il item is welded or the maximum radial thickness of the i

item exclusive of integral shell butt welding projec, corresponding to the thickness of the material, i.e.g (LST - T or) 2 A. This exemption does not exerr pt j#

A tions; (d) for flat heads, tubesheets, or flanges, use nhe either the weld metal (NC-2340),or the weldipg maximum shell thickness associated with the butt procedure qualifiestion (NC-4335) from impset tett-welding hub; ing.

l (c) for integral fittings used to attach process

'fI piping to the containment vessel or a containment

• Tgoy-temprature at oc shee the mit ductility transition 8

vessel nozzle, use the 1stger nominal thickness of the temperature NDT (ASTkt E 205). T.sor is 107 (6*C) belo. the

'""P"aium at *hich at leau t=0 tredracas Show aoM.

pipe connections.

rf I[o.ormance.eu senice Tempnaiure (t.sT) in the minimum temperature (1) bolting. inclu ing studs, nuts, and bolts, with mm) and less; of the fluid retained by the ecmponent or altematiuly the a nominal size of 1 in. ( 'nal cross sections! area of I

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+ A-212A D = FROM DWIT DATA

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FIG. B.7 NDT VALUES FOR " MILD STEELS "

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' f UNITED STATES gP:

NUCLEAR REGULATORY COMMISSION P

' WASHINGTON.D.c.20sss

% ug February 28.1986 0FFICs OF THE couwssionen

'iEl10RANDUM FOR:

_ Thomas A. Rehm Assister.t for Operations Office of the Executive Director for Operations FROM:

John H. Austin, Technical Assistert I

to Commissioner Asselst'ne

SUBJECT:

L!!1ERICK COMPLIANCE WITH 00C-51 Joe Halapatz, a former nenber of the Paterials Engineering Branch (MTEB),

brought the attached r.enoranda to the attention of Comissioner Asselstine.

The first memorandur r ated September 18, 1984, is the flTEB safety i

evaluation of the Lincrick 1 ar.d 2 containment pressure beundary materials.

Therein,l!TEB states:

"We have concluded that the results of our evaluation and the augmented inservice inspection progran for these valves will provide

.rev.crable assurance of compliance with the requirements of GDC 51, contingent on confinnaticn by the augmented ISI that the 'brinkage flaws (mistire in the valve bodies on entering service have not prcpagated to either of the surfaces.

Struld the augmented ISI oisclose that these flaws brac prepacated to either of the surfaces, the valves ther are te he replaced by the licensee."

The seccrd tr.cmorandum, dated October 4,1984, is frer i frerick to NRR and con'irms an October 1.19P4 telephone conversation between the applicant and e nenber of MTEB. The latter memorandum ccrtrirs the following cornitnent:

The augmented inservice inspection peccrair "will include inspection of the entire surface of the valve Fedics, both internal and external, by surface or other examination techr'ct:e receotable to the staf'. Tho, inspection program will, after appropriate surface preparation, be of sufficient sensitivity, to detect t. r rimum crack length of 31 4

inches."

While this comitment was apparent'," screwhat modified in an Octrber le.,

1984 letter from the applicant (alse atteched), Mr. Halapatz is concerred i

that there is a substantial techricel difference between the official branch position in the SER as quoted abevr. erd the positions stated in the October 4 and 12,1984 letterr. thet were developed in inforwal telrphone conversations with NRR individuals.

Thrt d"ference and the way the

" understandings" were apparnetly develnped, Mr. Halapat: claims, creates a g7L

f;s e o

significant safety issue at Limerick and ruhverts tho safete res4ow procedures of the Agency.

I wnuld appreciate learning what'the official MTEB position is regarding cracks in the valves discussed in the Septerter 18, 1084 memorandum, including the scnsitivity of the required inspecticns of the valves for surface cracks and actions that nust be taken if any curface cracks are detected. A response by March 21, 1986 would he appreciated if that would net adversely impact onocirc work.

Attachnents-As stated I

1 s

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h. [ [

umTso:TATES NUCLEAR REGULATORY COMMISSION 3

s memworos.o. c. noses MAR 211986 Docket Nos.: 50-352 50-353 MEMORANDUM FOR: John H. Austin, Technical Assistant to Connissioner Asselstine FROM:

T. A. Rehm Assistant for Operations, EDO

SUBJECT:

LIMERICK COMPLIANCE WITH GDC-51 In your memorandum to me dated February 28, 1986, you requested certain information with respect to a particular staff input for the Limerick Safety Evaluation Report and several subsequent letters from the Philadelphia Electric Company (PECo) in 1984. The documents deal with potential shrinkage flaws in feedwater check valve bodies. We believe that this issue was appropriately resolved and that there was no compromise of the staff's safety review procedures. The following discussion which includes some pertinent background information, will serve to place this issue in the proper perspective, and should provide the information you requested.

During the staff's review of the Final Safety Analysis Report for the Limerick Generating Station in 1983 and 1984, the staff determined that two cast bodies for the feedwater check valves contained shrinkage cracks that could compromise the valves' conformance with GDC-51 if the cracks were to propagate while in service. Such propagation of the shrinkage cracks has also occurred in the past and has been documented for similar valves at other plants. After a fracture mechanics evaluation was com-pleted and subm.itted by PEco for the staff's review, an independent fracture mechanics evaluation was performed by the NRC staff. It was the staff that a special inservice inspection of these determined by(valve 1F074 A and 8) would be required.

valve bodies The resolution of this issue is addressed in the NRC staff's input to the SER, dated September 18, 1984, two PEco letters dated October 4 and 12, 1984, and Supplement No. 3 to the Limerick SER (NUREG-0991) issued in October 1984. Specifically, the supplement to the SER stated that:

"We have concluded from these calculations that although the valve body may not quite meet the Appendix G requirements, even our most margin against failure (at least a factor of two) gin.

conservative approach still shows some safety mar Adequate will exist under the most probable loading conditions.

Contact:

R. E. Martin x29472

% h S-% h S f

u s..,

. These conclusions assume that the flaw size assumed will not be exceeded significantly. However, service experience on similar castings has disclosed that normal, acceptable shrinkage may be extended by cracking during service.

We, therefore, recommended to the applicant that these valves be inspected for surface cracks on the inside and outside surfaces at the first refueling outage and at other times when the valve is disassembled for maintenance.

The applicant has committed, by letter dated October 12, 1984, to including this augmented inspection by surface examination or other methods acceptable to the staff, which will be determined during the staff's review of the inservice inspection program.

We have concluded that the results of our evaluation and the augmented inservice inspection program for these valves will provide reasonable assurance of compliance with the requirements of GDC-51.

It will be confirmed by the augmented ISI that the shrinkage flaws existing in the valve bodies on entering service have not propagated to either of the surfaces. Should the augmented ISI disclose that these flaws have propagated to either of the surfaces, the valves are then to be replaced by the Itcensee."

With respect to the issue raised by the PECo letter of October 4,1984, i.e., whether a surface crack of up to 3.5 inches in length would be acceptable, the staff's position has been and continues to be that if the valve body inspections disclose that internal shrinkage cracks have propagated to either surface, the valves are to be' replaced, regardless of the length of the surface indication.

It should be noted that the alternative approach proposed by PECo in its letter of October 4,1984, was found to be unacceptable by the staff and PECo was so informed; whereupon, the PECo proposed approach described in its letter dated October 12, 1984, was submitted and determined to be acceptable. The staff findings reflected in the September 18, 1984, staff memorandum referred to in your memorandum were included in SSER-3, as discussed above. There has been no change in the staff's position from that stated in SSER-3, Assistant for Operations, E00 cc:

P. Polk C. Ader J. Myer M. Clausen OPE OGC SECY