ML19347A871

Fracture Toughness of Steam Generator & Reactor Coolant Pump Supports,Arkansas Unit 2, Interim Technical Evaluation Rept

ML19347A871
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/30/1980
From:	Allten A, Noell P, Stilwell T FRANKLIN INSTITUTE
To:	Fair J Office of Nuclear Reactor Regulation
Shared Package
ML19347A866	List: ML19347A865 ML19347A871
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-160, NUDOCS 8009300444
Download: ML19347A871 (15)

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TECHNICAL EVALUATION REPORT FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS ALKANSAS POWER AND LIGHT COMPANY i ARKANSAS NUCLEAR ONE UNIT 2 m_

NRC DOCKET NO. 50-368 NRC TAC NO. 12066 FRC PROJECT C5257 NRC CONTRACT NO. NRC 03-79-118 FRC TASK 160 Preparedby Franklin Research Center Authors: T.C.Stilwell, P.N.Noell, The Parkway at Twentieth Street A.G. Allten, K.E.Dorschu Philadelphia, PA 19103 FRC Group Leader: T. c. Stilwell Preparedfor Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: J. R. Fair September, 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 agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any 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 wouti not infringe privately owned rights, al

00. Franklin Research Center A Division of The Franklin Institute The Bernarren Frankhn Parkway. PNia.. Pa. 19103(2151448 1000 8009300444

(INTERIM)

TECHNICAL EVALUATION REPORT FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS '

ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE UNIT 2 1

NRC DOCKET NO. 50-368 NRC TAC NO. 12066 FRC PROJECT C5257 NRC CONTRACT NO. NRC-03 79118 FRC TASK 160 l

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Preparedby Franklin Research Center Authors: T.C.stilwell, P.N.Noe11, The Parkway at Twentieth Street A.G.Allten, K.E.Dorschu Philadelphia, PA 19103 FRC Group Leader: T. c. stilwell Preparedfor Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: J. R. Fair september, 1980 This report was prepared as an aceunt of work sponsored by an c;ency of the United States Government. Neither the United States Government nor any agency thereOf, or any of their employees, makes any warranty, expressed or Impiled, or assumes any legal liability or responsibility for any third party's use, or the results of l 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 rights.

A 1 Franklin Research Center A Division of The Franklin Institute The Benpamin Frankhn Parkway. Pheia., Pa t9 403 (2151448-1000

. TER-C5257-160 (Intoris)

CONTENTS Section Title h 1

SUMMARY

. . . . . . . . . . . . . 1 2 INTRODUCTION . . . . . . . . . . . . 2 3 BACKGROUND . . . . . . .. . . . . . 2 4 CRIT'!RIA APPLIED IN TR EVALUATION . . . . . . . 5 4.1 Fracture-Toughness Grouping of Materials Used in Support Construction . . . . . . . 5 4.2 Plant Grouping for Fracture-Toughness Ranking of S/G and RCP Support Structures . . . 5 4.3 Criteria for Fracture-Toughness Adequacy of S/G and RCP Supports . . . . . . . . . 6 5 TECHNICAL EVALUATION . . . . . . . . . . 7 5.1 Use of Group I Materials . . . . . . . . 9 5.2 Lack of Information, Supports . . . . . . 10 5.3 Lack of Information, Welding . . . . . . . 10 6 CONCLUSION . . . . . . . . . . . . . 11 TABLE Section Title Pg 5.1 COMPONENI SUPPORT

SUMMARY

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FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS

1.

SUMMARY

Information concerning aspects of the fracture-toughness design of the steam generator (S/G) and reactor coolant pump (RCP) supports for the Arkansas One Unit 2 nucisar power station was submitted to the Director of Nuclear Regulation by the Arkansas Power and Light Company (APL) letter dated December 19, 1978. This information was reviewed at the Franklin Research h eer (FRC) and evaluated in accordance with the criteria of the Nuclear Regulatory Commission (NRC) as set forth in NUREG 0577-Draft (henceforth referred to simply as NUREG 0577).

FRC found that steel, ordered to the American Saciety for Testing and Materials (ASTM) Specification A283 is used in lugs critical to the horizontal seismic restraint of the RCPs. This steel is classified as a Group I (rela-tively poor fracture toughness) mata: rial by NUREG 0577 and fails to meet the NUREG 0577 nil ductility tempe:ature (NDT) screening criteria.

In addition, FRC found that the fracture-toughness adequacy of certain support components had not been demonstrated because the submittal either omitted or furnished incomplete information needed for definitive evaluation.

These components include: (1) the S/G snubber system support anchorage and (2) veldsents used in the S/G supports.

l It is possible that ASTM A283 steel can be shown to have adequace fracture I toughness for this particular application and that the fracture-toughness adequacy of all components in the other systems listed above can subsequently be demonstrated.

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l TER-C5257-160 (Interim) 4 Pending further clarification of these principal concerns, FRC recosmends that a tentative Group I plant classification for fracture toughness of S/G and RCP supports be assigned to Arkansas one Unit 2.

2. INTRODUCTION This report provides a technical evaluation of information supplied by APL with its letter of December 19, 1978, to the Director of Nuclear Reactor Regu-lation. The information concerns the fracture-toughness design of supports for the S/G and RCPs for Arkansas One Unit 2 Le objective of the evaluation is to rank the design for fracture-toughre so istegrity on a relative scale in accordan:e with the grouping scheme acI criteria established in NUREG 0577.

The ranking is considered tentative because:

1. It is based solely on review of the information submitted.

2. NUREG 0577 and the criteria it contains had not been developed 1 at the time the information was requested from the licensee.

Additional, more specific information relevant to the plant grouping may be on hand and might have been submitted had NUREG 0577 been available to provide guidance at the time '

the information was solicited.

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3. BACKGROUND l During the course of the NRC licensing review for two pressurized water

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reactors (PWR), North Anna Units 1 and 2, questions were raised regarding the l fracture-toughness adequacy of certain members of the S/G and RCP supports.

The potential for lamellar tearing in some support members was also questioned.

The staff's concern in the North Anna licensing process was that not enough attention might have been paid to the selection of materials for, and fabrication of, the S/G and RCP supports.

Fracture toughness of a material is a measure of its capability to absorb energy without failure or damage. Generally, a material is considered " tough" nklin Research Center i A D>sson cd The Frenhen inneaute

TER-C5257-160 (Interim) l when, under stated conditions of stress and temperature, the material can withstand loading to its design 1 Lait in the presence of flaws. Toughness also implies that under specified conditions the material has the capability to arrest the growth of a flaw. A lack of adequate toughness (accompanied by the combination of low operating temperature, presence of flaws, and nonre-dundancy of critical support members) could result in failure of the support structure under postulated accident conditions, specifically, loss-of-coolant 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 reicvant ASTM specifications. These tests indicated that material used in certain support members has 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 support beams in question to a minimum of 225 F every time, throughout the life of the plant, the reactor coolant system (RCS) is pressurizei above 1000 psig. The NRC staff found this to be an acceptable resolution.

Because similar materials 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 Activity A-12--Potential for Low Practure Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports."

Since the original licensing action (North Anna Units 1 and 2) involved only the S/G and RCP supports of' PWRs, the staff's initial efforts were directed toward examination of the corresponding supports at other PWR facilities. However, the staff has kept in mind the possibility of ex-pandlug its review to include other support structures in PWR plants and support structures in boiling water reactor (BWR) plaats.

The integrity of support embedmonts was not questioned during the North Anna licensing action, and

• nhasis was consequently placed on resolving the most immediate generic issue--whether or act problems similar to those un-4 Obuacm..

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TER-C5257-160 (Interim) covered at North Anna exist at other facilities. It was the staff's judgment that inclusion of an evaluation of support embedments in the initial review would require detailed, plant-specific investigations that were beyond the scope of the preliminary, overall generic review. Such censiderations were deemed more suited to a subsequent phase when more detailed investigations of individual plants might be undertaken.

Requests for information were sent to licensees in late 1977. Responses to these requests were received during 1978.

Sansia Laboratories of Albuquerque, New Mexico, was retained to assist the staff in the review and analysis of the information received from licensees and applicants. Based on an analysis of the information, the technical j studies made by Sandia Laboratories, and review of the issue by the NRC staff, the NRC developed an NRC staff technical position on these issues. This is presented in NUREG 0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant 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 criteria to the support structures of a number of PWR plants to achieve plant groupings according to the relative fracture-coughness integrity of these supports. The plant ratings are:

e Group I (lowest) e Group II (intermediate) e Group III (highest)

NUREG 0577 also emphasizes the tentative character of these rankings, acknowledging that a number of plants were classified as Group I because licensees had not submitted all the information needed for definitive classification. Therefore, they had not demonstrated that their plant merited a higher ranking. In this regard NUREG 0577 states:

Receipt of such (i.e., currently unsubmitted) information could result in the plants being moved to a lower susceptibility (to brittle failure) group after very little additional analysis.

Several power plants, Arkansas One Unit 2 among them, were not reviewed during the generic study and, therefore, received no group ranking.

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This technical evaluation applies the criteria of NUREC 0577 to the S/G and RCP supports for Arkansas One Unit 2 in order io provide a preliminary assessment ot the fracture-toughness adequacy of these supports leading to a plant rt.nking.

4. CRITERIA APPLIED IN THE EVALUATION 4.1 FRACTURE-TOUGHNESS GROUPING OF MATERIALS USED IN SUPPORT CONSTRUCTION 4.1.1 Criterion Table 4.6-Material Groups-of Appendix C to NUREG 0577 groups materials according to their relative fracture toughness ass e Group I (poorest) e Group II (intermediate) e Group III (best) 4.1.2 Interpretation ,

If no supplementary requirements were called out in the material specifi-cation ai: sed at procuring a product with fracture-toughness properties superior to those routinely supplied under the ASTM (or other standard) 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, consideration was given to crediting this specific material order with an improved material-group rating. ,

4.2 PLANT GROUPING FOR TRACT'JRE-TOUGHNESS RANKING OF S/G AND RCP SUPPORT STRUCTURES 4.2.1 Criterion Plants are classified on the basis of the construction materials used in the support 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.)

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Plants were assigned a plant-group ranking identical to the material-group ranking of the least fracture-cough material used in the construction, provided this usage is important to support integrity.

4.3 CRITERIA 70R FRACTURE-TOUCHNESS ADEQUACY OF S/G AND RCP SUPPORTS It is the clear intent of NUREG 0577 that licensees demonstrate the j fracture-toughness adequacy of the S/G and ACP supports or that they take appropriate corrective measures to assure their fracture-toughness integrity.

NUREG 0577 provides guidance for such demonstrations.

4.3.1 NDT Criteria for Screening

, 300F  :

NDT + 1.30 + or <T i 600F 8uPPorts (OF) wherus e NDT is the mean nil ductility transition temperature appropriate to the material as given by Table 4.4 of Appendix C to NUREG 0577.

a.

e a is the standard deviation for the data used to determine NDT as listed in Table 4.4.

e Tsupport is the lowest metal temperature that the support member will ever experience throughout the plant life when the plant is in an operational state. In the absence of measured, plant-specific l data, Tsupport is taken as 750F.

e The temperature term, 300F or 600F, is an allowance for section size (300F for thin sections and 600F for thick sections).

l 4.3.2 Interpretation 1

If evidence is furnished by the licenaee proving that other values of NDT, o, or T,,pp,,g, are actually valid for S/G or RCP supports and materials in tha licensee's plant, such data may be used. In the absence of acceptable, contrary evidence, values are used as stipulated above.

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NUREG 0577 also recognises that fracture-toughness integrity is a compler matter involving a number of interrelated factors, most of which are plant-  !

1 specific. Consequently, demonstration of compliance with the screening criteria is but one means of providing satisfactory assurance of fracture-l toughness adequacy.

l NUREG 0577 not only recognizes that other means of showing compliance with the intent of NUREG 0577 are possible, but also offers extensive guidance i relating to several approaches by which such a demonstration any be achieved.

Because of the plant-specific character that such demonstrations mList take, NUREG-0577 does not restrict the licensees to any single approach but, instead, encourages each licensee to review the fracture-toughness adequacy of his S/G and RCP supports and submit ev! dance of his findings.

5. TECHNICAL EVALUATION A review has been made of information contained in the response dated December 19, 1978, from APL's Mr. D. H. Williams to a request for information, dated December 5,1977, from Mr. J. F. Stolz of the NRC. The request sought information concerning the fracture toughness of, and the potential for lamellar tearing in, the S/G and RCP supports. A copy of this request (in generic form) may be found in NUREG 0577, Appendix B. Key items from APL's response concerning Arkansas one Unit 2 were condensed to tabular form re d are presented as Table 5.1.

FRC's review addressed only the fracture-toughness issues. Issues related to the potential for lamellar tearing will be assessed after results of an ongoing basic study of the problem becces available.

Considerabia effort was made by APL to document material properties with mill test and inspectior. rsports. However, asny of tha reports furnished re-late to S/G and RCP components and not to the support structures themselves.

The available information indicates that, in general, choice of materials was made with care. Nevertheless, FRC's review disclosed several areas where a

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TABLE 5.1 ,

COMPONENT SUPPORT SUMMART FIANTs ARKANCAS NUCLEAR 090E UNIT 2 e:

S IITILITT NSSS AE SUPPORT SUPPLIER

> yg #k' y Arkanese Power C.E. Bechtel CE 3 see Light Co.

E e NATERI AI.S MAXIMUM ALIAtiASLE DESIGN STRESS

'h 7 Mll.1 CERTS. NEAT NDE 000 FRACTU-TOUCNNESS IEMBRANE & TNROUCN h

33 TYPE AVAllABIE

• TREATTEleT MATERIAL TEST BENDING (IIORMAL) TIIICRIESS k SA516 CR 70 Ch., Tee., CS IIore. + Temper UT CVN SA516, Faulted -

54540 CLI CR B21 Ch., Ten., CS Tee UT, MP CVN 60.66 kai (1.8 Sy )

SA533 CLI CR S 03., Ten., CS Nora.

• Temper UT, MP CVN, 70 ft Ib aug SA333 Faulted:

9 10'F S/C Skirt 54.3 kai 9/C Skirt SA216 CE WCC Ch., Ten. More. + Temper trT, MP -

A283 Isot fcund - - -

SA216, Faulted:

28.7 kai (0.6 Sy )

SA240 Type 304 - - - - SA487, Faulted A4340 (Degassed) Ch., JT Isore.

• Q & Temper + S.R. - -

31.5 (0.35 S.)

, A193 57 Cr 4140 Tes Yes - -

SA351, Emergency 3 cn 29 kei (1.5 S 1/

8 SA487 4N Ten 18ers. + Temper

• S.R. trf, MP -

34.8 koi (l.8 S )

SA351 CFSM - -

RT

• Ch. = f:hemistryg Ten.

• Tenellet CS = Crain sizeg JT = Joeiny teett S.R. = Stress Relief.

FARRICATION METNODS USED TO IIDE AND WELDillC IdELDIMC POST-WELDilIC PREVENT lAMELlAR INSPECTIONS PROCESS PROCEDURE TREAT 1ENT TEARINC PERFORMED 10 0 heavy weldsento -

in structure DESICM TYPE OF SUPPORT OGDE USED thADING 00tIDIT10MS MINIMUM TEMPERATURE OF SUPPORT Sliding Pedestal, Skirt Support S/G&RCP - Normat = Dtff*T* Pres.

S/C&RCP - Upset = Isoemal + l/2 SSE Ambient in support area is "about 850F."

Ck D L8 S/CERCP - Emergency = Normal + SSE $$N

• 1 S/C - Faulted = Normal + SSE

• 12CA F I o

TER-C5257-160 (Interim) fracture-toughness adequacy has not been demonstrated. These, with one ex-ception, involve incomplete or missing information. Items of concern are susssarized briefly below:

Use of Group I Materials. ASTM A283 steel, classified by NUREG 0577 as j

Group I (relatively poor fracture toughness), is used for a lug attachment in i the RCP horizontal snubber system.

4 Iack of Information, supports. No information was found in APL's f response on materials or design details of the anchorage of the S/G snubber system to the building walls. Similar concerns relating to the RCP vertical hangers and horizontal snubbing devices have been resolved by recent 4

information from the NRC.

1 Lack of Stress Data, RCP Supports. No calculated or allowable stresses 1

were given for the RCP vertical hangers and hanger lugs for faulted loading conditions.

Lack of Re2d Information. Little information was furnished on weld materials, welding procedures, and inspection, methods for the S/G and RCP supports.

The above items of concern are discussed in more detail in the following i sections.

5.1 USE OF GROUP I MATERIALS Table I of the APL response letter Dectober 19, 1978, lists ASTM A283 steel as the lug material for the RCP horizontal seismic snubber mechanism.

These lugs are welded to the RCP actor housing and appear on General Electric Company Dwg. 816E310 Rev. 5.

Lug dimensions sre not shown, but the lugs appear to be constracted from heavy A283 plate. No designated grade of A283 could be identified from the drawings, nor was other reference to the material found (except for tabulation in Table I of APL's response.)

ASTM A283 is classified as a Group I steel by NUREG 0577. On the basis of NUREG 0577's screening criteria, A283 is of questionable fracture toughness

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'99 TER-C5257-160 (Interim) even at APL's estimated minimum ambient temperature of 85 F.

The lugs are clearly important to structural integrity. Postulated fracture of the RCP enubber lugs under faulted loading conditions might transfer heavy loads to the RCP vertical support hangers.

It may, however, he possible to demonstrate the acceptability of this particular application of A283 by showing that alternative criteria, as provided by NUREG 0577 for fracture-toughness adequacy, are met.

5.2 LACK OF INFORMATION, SUPPORTS )

l Item 1 of APL's response describes the supports of the RCPs as follows:  ;

i "The reactor coolant pumps for ANO-2 are supported by four vertical spring hangers and two horizontal hydraulic snubbers 4

which are attached between legs welded to the pump assembly and wall brackets attached to the building structure. The pump supports are sized to carry the dead weight and seismic loads of the pump."

No information or drawings describing the material, design dimensions, anchorage, or stresses for the RCP vertical spring hanger supports were found in the documents furnished. However, the NRC was able to supply additional drawings showing RCP spring hangers, horizontal snubbers, and supplementary RCP restraint by wire rope. It is FRC's understanding that this design had been subjected to a recent review by others and found .idequate.

Although drawings were included showing the lateral support of the steam generator by a snubber system, these drawings fail to show how this system is bolted to the building wall and the asterials used in this anchorage.

5.3 LACK OF INFORMATION, WELDING Little information was found in the APL letter of response concerning weldsents. I Combustion Engineering's drawing No. E234-803, entitled Skirt De. tails and Assembly, of the S/G Support Skirt shows that this conical skirt is constructed of SA533 Grade B steel, and welded:

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TER-C5257-160 (Interim) a) At its base to a massive support skirt flange b) At its top to the supported structure c) Longitudinally in four places to unite the four conical segments which constitute the ekirt.

CE's drawing E234-804, External Supports, also shows the snubber lug assembly to be a weldment and that these lugs are, in turn, welded to the S/G upper shell.

No information was furnished concerning welding consumables, welding procedures, or weld inspection for any of these welds. It was, however, suggested that such information might be inspected at Combustion Engineering's office.

The only other major welding occurs in the attachment of the reactor coolant pump support lugs. These are fabricated using stainless steel and are welding to the pump casing.

Although no welding procedures are provided for these, the available records indicate that the stainless steel weld metals contain sufficient ferrite levels to prevent weld metal cracking. Since austenitic stainless l steels are reasonably tough, the procedures are less critical than those  !

l used to weld many carbon and low alloy steels for these applications. i However, the specific materials and procedures for all welds need to be 1

reviewed to render a more accura?.e judgment.  ;

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6. CONCLUSION Information received from APL relating to the fracture-toughness design of the S/G and RCP support systems for Arkansas One Unit 2 was reviewed at 4

FRC and evaluated in tecordance with the criteria of NUREG 0577.

It was found that lugs for the RCP seismic snubber system are made of ASTM A283 steel. This steel is classified as a Group I material by NUREG 0577, and may exhibit low fracture toughness at the 85 F estimated temperature near the

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RCP. No fracture toughness data or other additional requirements concerning the A283 were found in the information submitted.

In addition, uncertainties stem from incomplete information. These omissions include: 1) description of materials and structural method used in anchorage of the S/G snubber systems to the buildird valls, and 2) information concerning details on velding procedure.

Based on the use of a Group I steel and because the infor'sation furnished was insufficient to demonstrate that the S/G and RCP supports merit more favor-able classification, it is recoimsended that Arkansas One Unit 2 be Statatively assigned to NUREG 0577 Plant Group I, pending clarification of items of con-cern.

It is possible that the A283 steel could be shown to possess adequate fracture toughness for its application in Arkansas One Unit 2, and that fracture-toughness integrity can also be demonstrated for components which could not be evaluated because of insufficient information.

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