ML20137E295

Safety Evaluation Supporting Amends 125 & 119 to Licenses NPF-2 & NPF-8,respectively

ML20137E295
Person / Time
Site:	Farley
Issue date:	03/24/1997
From:	NRC (Affiliation Not Assigned)
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ML20137E281	List: ML20137E276 ML20137E289 ML20137E295
References
NUDOCS 9703270148
Download: ML20137E295 (8)

Text

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. ensam anni a

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 125 TO FACILITY OPERATING LICENSE NO. NPF-2  !

AND AMENDMENT NO. 119TO FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 4

DOCKET NOS. 50-348 AND 50-364  !

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1.0 INTRODUCTION

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By letter dated January 10, 1997, as supplemented by letter dated February 24, <

1997, the Southern Nuclear Operating Company, Inc., et al. (the licensee), ,

submitted a request for changes to the Joseph M. Farley Nuclear Plant, Units 1 i and 2, Technical Specifications (TS). The requested changes would incorporate the latest revised topical reports governing the installation of laser welded steam generator tube sleeves designed by Westinghouse (W). The licensee is l

presently licensed to use Westinghouse sleeves for repairing steam generator tubes as an alternative to plugging. The proposed changes update the existing Farley plant Technical Specifications to incorporate the latest advances in Westinghouse sleeve installation technology and add certain current industry recommended practices for sleeved tube reinspections. In addition, the reference to a one-cycle implementation of L*, which expired at the last Unit 2 outage would be deleted from the Unit 2 TSs. The February 24, 1997, letter provided clarifying information that did not change the original application and the initial proposed no significant hazards consideration determination published in the Federal Reaister on January 29, 1997 (61 FR 4355).

The revised topical reports detail the latest design and installation features i

of three types of steam generator tube sleeves: a full depth tubesheet sleeve, an elevated tubesheet sleeve, and a tube support plate sleeve. The updated design for Westinghouse sleeves is to laser weld and heat treat the freespan joint (s) and mechanically roll the tubesheet joints. As previously licensed, the Westinghouse sleeves for use at the Farley plant specified a

tubesheet joint incorporating both a rolled and welded joint, and a freespan welded joint with an optional heat treatment.

Additionally, due to certain construction details applicable to the Farley units, the proposed change also modifies the normal rolled joint to incorporate a double roll.

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The proposed changes are based upon the experiences gained and improvements incorporated by Westinghouse during the last 2 years during large sleeving campaigns at other facilities.

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1 l  ! Extensive analyses and testing were performed on the design and process j

modifications to demonstrate that Regulatory and Code design criteria for sleeved tubes were satisfied under normal operating and postulated accident conditions. The details of the sleeve qualifications for Farley are discussed in Westinghouse reports WCAP-13088, " Westinghouse Series 44 and 51 Steam Generator Generic Sleeving Report, Laser Welded Sleeves" (proprietary), dated January 1997, and WCAP-14740, " Specific Application of Laser Welded Sleeves for the Farley Units 1 and 2 Steam Generators" (proprietary), dated January 1997.

The staff has previously reviewed similar W documents supporting TS amendments for sleeve installations at other plants. The bulk of the technical and

. regulatory issues for the present request are identical to those reviewed in previous Safety Evaluations (SEs) concerning the use of M laser welded sleeves. This SE summarizes the principal issues discussed in previous reviews and adds a discussion of those warranting revision, amplification,

, or inclusion based upon current experience. Details of the prior staff evaluation of W sleeves may be found in SEs for Calvert Cliffs Nuclear Power

Plant Units 1 and 2, Docket Nos. 50-317 and 50-318, dated March 22, 1996; DC 1 Cook Nuclear Power Plant Unit 1 Docket No. 50-315, dated January 4,1996; and Maine Yankee Nuclear Power Plant, Docket No. 50-309, dated May 22, 1995.

Additionally, prior evaluations of Westinghouse sleeves have been performed for the Farley plant. The relevant TS amendments were dated September 18, l 1987, October 22, 1990, and November 20, 1996.

2.0 SUtWARY OF PREVIOUS REVIEWS Previous staff evaluations of W sleeves addressed the technical adequacy of i the sleeves in the principal areas of pressure retaining component design:

structural requirements, material of construction, welding and post weld heat i treatment eifects, and nondestructive examination. Along with these design evaluations, the staff has included evaluations of sleeve design changes based upon operating experiences with previous sleeving installations. The staff position and findings regarding sleeving methods are summarized below:

2.1 Structural Requirements

. The function of sleeves is to restore the structural integrity of the tube I pressure boundary. Consequently, structural analyses were performed for a i

variety of loadings including design pressure, operating transients, and other parameters selected to envelop loads imposed during normal operating, upset, and accident conditions. Stress analyses of sleeved tube assemblies were

performed in accordance with the requirements of the ASME Boiler and Pressure

! Vessel Code,Section III. These analyses, along with the results of qualification testin{. and previous plant operating experience, were cited to demonstrate the sleeved tube assembly is capable of restoring steam generator tube structural integrity.

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2.2 Material of Construction 1

l The sleeves are fabricated from thermally treated alloy 690, 'a Code-approved material (ASME $B-163) covered by ASME Code Case N-20. The staff found the use of alloy 690 is an improvement over the alloy 600 material used in the

original SG tubing. Corrosion tests conducted under Electric Power Research i Institute (EPRI) sponsorship confirmed test results regarding the improved corrosion resistance of alloy 690 over that of alloy 600. Accelerated stress corrosion tests in caustic and aqueous chloride solutions also indicated alloy i 690 resists general corrosion in aggressive environments. Isothermal tests in j high purity water have shown that, at normal stress levels, alloy 690 has high i resistance to intergranular stress corrosion cracking (IGSCC) in extended high

temperature exposure. The NRC staff concluded, as a result of these i laboratory corrosion tests, that alloy 690 is acceptable for use in nuclear i power plants. The NRC endorsed the use of Code Case N-20 in Regulatory Guide j j 1.85, " Materials Code Case Acceptability, ASME Section III, Division 1." The i NRC staff has approved the use of alloy 690 tubing in replacement steam )

l generators as well as sleeving applications.

1 2.3 Welding and Post Weld Heat Treatment l Automatic autogenous laser welding is employed to join the sleeve to the  ;

i parent tube in the free span regions. The application of this process to the i j W sleeve design was specifically qualified and demonstrated during laboratory i

tests employing full scale sleeve / tube mockups. Qualification of the welding l

} procedures and welding equipment operators was performed in accordance with '

j the requirements of the ASME Code,Section IX.

I Accelerated corrosion tests have confirmed that a post weld heat treatment (PWHT) significantly improves the IGSCC resistance of the alloy 600 parent tube material in the weld zone. A PWHT reduces the residual stresses i resulting from welding. Residual stresses from forming operations (such as

bending, welding, etc.) are known to be a principal contributor to IGSCC in i alloy 600. Performance of a PWHT greatly reduces the residual stresses from i- welding thereby enhancing the IGSCC resistance of the alloy 600 portion of the

j. weld zone. The alloy 690 sleeve material is highly resistant to IGSCC either

with or without PWHT. All free span laser welded joints will be heat treated l in accordance with the M generic sleeving report (WCAP-13088) and the NRR

staff conclusion that PWHT enhances weld joint resistance to IGSCC.

The rolled joint used to join the sleeve to the tube within the tubesheet i effectively isolates the alloy 600 of the parent tube from the environment and, i thus, is not susceptible to IGSCC. Stress relief of these joints is i unwarranted. PWHT of lower joint seal welds (where used) is undecirable due

to potentially deleterious effects upon the tubesheet material and the

! integrity of the rolled joint.

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2.4 PWHT and Tube Lockup +

i Field 3xperience with the installation of welded sleeves with PWHT indicated i

that steam generator tubes may be constrained in their tube supports (" tube  ;

j lockup"). The principal result of such tube locking is a residual stress i ("far-field" stress) remaining after the heat treatment is completed. Full l scale laboratory mock-ups of sleeved tube assemblies allowed study of the contributing factors to the far-field stress. Based upon these laboratory .

test, the sleeve installation procedure was modified to minimize the far- l l field stress. Strain gage measurements of the remaining residual stress have ,

shown it to be moderate compared to the stress resulting from welding without subsequent PWHT.

l The effectiveness of the PWHT with modified installation procedure for i minimizing the far-field stress has been further verified with accelerated i corrosion tests. W has developed a proprietary accelerated corrosion test

! that is used to rank the relative IGSCC resistance of tubes, welds or other i components. Welded tube / sleeve samples with and without PWHT and subjected to

operating pressure plus far-field stresses have been tested for susceptibility i to IGSCC. These test results have consistently and clearly demonstrated the i superiority of the specimens with PWHT over all others.

l 2.5 Service Life Predictions for Sleeved Steam Generator Tubes i

i The staff position on sleeving considers the method unable to assure an

! unlimited service life for a repaired tube. The conservative view is that

sleeving creates new locations in the parent tube, which may be susceptible ,

to IGSCC after new incubation times are expended. Incubation times are not  !

quantified. They are observed to vary between individual steam generators and l

the various tubes within, based upon prior experiences with U-bend and roll j transition cracking.  ;

i l- This staff position that sleeving has limited service life is based upon the j

circumstances of the sleeving processes. Sleeve installation methods can '

j enhance one or two of the conditions necessary for IGSCC. The primary contributor is the residual stress resulting from the various joining methods.

In addition, the local environment of the tube may be altered as a result of i the formation of a wetted crevice between the tube and sleeve. Remediation of these contributors would benefit sleeved tube life. Of the two, stress

. relieving may be the most beneficial given the underlying causes of IGSCC and

! present sleeve designs. As discussed earlier, the sleeve installation j procedure includes a PWHT of the weld joints to increase the resistance to

IGSCC.

I 2.6 Nondestructive Examination 1 The sleeve assemblies can be inspected by nondestructive techniques in

accordance with the recommendations of Regulatory Guide 1.83, " Inservice

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Inspection of Pressurized Water Reactor Steam Generator Tubes."

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l Nondestructive examination of sleeved tubes is conducted in two primary ways.

! Ultrasonic testing (UT) is performed after welding to confirm the laser welds

' are consistent w'th critical process dimensions and are of acceptable weld j quality. W presented data on a UT system demonstrating that post weld

- -examinations of the sleeve / tube assembly will be adequate. Standards that

! included undersized welds were used in the qualification of the UT technique.

l The results of the qualification tests demonstrate the system can confirm that i there is a continuous metallurgical bond between the sleeve and tube and the l weld size (width) meets minimum acceptable dimensions.

i Eddy current testing (ECT) is then used to establish baseline inspection data

! for every installed sleeve / tube. This data is compared with subsequent ECT

! inspections to aid in identifying any possible service induced degradation,

, should it occur. In performing these inspections, the licensee will use

! Electric Power Research Institute (EPRI) "PWR Steam Generator Tube Examination l Guidelines" Appendix G qualified personnel and Appendix H qualified ECT

techniques. For future sleeve / tube inspections, the licensee committed to following the most current revision of the EPRI guidelines in terms of inspection scope and expansion criteria as reflected in adoption of TS Table 4.4-3, " Steam Generator Repaired Tube Inspection."

3.0 DISCUSSION The previous section addressed generic topics applicable to steam generator tube sleeve installations using W 1aser welded sleeves. For the Farley amendment request, plant specific modifications were also proposed.

3.1 Modified Sleeve Rolling Procedure for Tubesheet Joints Due to certain original construction details peculiar to Farley Unit 1, the licensee requested a modification of the standard W lower rolled joint for use.at the Farley facility. The construction difference between the two units' steam generators involves the WEXTEX expansion method that was employed on Unit 1. Unit 2's steam generators were constructed with full depth rolling. In the interests of commonality, the licensee sought a lower joint rolling method that would be applicable to both units. Consequently, the licensee engaged Westinghouse to develop a modified rolling procedure (called a two-roll pass lower joint) for use at Farley.

Since the modified rolling procedure was a departure from that previously approved for other installations, a new series of qualification tests were performed as detailed in WCAP-14740. The principal tests concern measured leak rate (if any) and structural integrity for all design conditions. Mock-ups of the modified rolled joint were produced and-laboratory tested for conformance with the requirements for leak rate and structural capability.

Leak test specimens subjected to a range of pressures (reflecting primary-to-secondary pressure differentials) showed no test samples with leak rates

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beyond a tiny fraction of a drop per minute. Using the average measured leak i

rate for all the tested samples and assuming a worst case accident condition (steam line break) with 1000 sleeves per generator would yield a resulting leak rate measurable in the crops per minute range. This is an insignificant fraction of the allowable primary-to-secondary accident-induced leakage limit.

For Farley, this limit is 11.4 gpm. (An accident-induced primary-to-secondary l leakage limit is determined to ensure that the 10 CFR Part 100 and General '

Design Criterion 19 radiation limits are met). Therefore, it can be concluded i that primary-to-secondary leakage under these worst case conditions would be insignificant or zero for plant normal and postulated faulted event pressure conditions for the two-step rolled joint configuration.

The mechanical strength of the two-step rolled joint was tested by loading

, mock-up joints to failure and noting the load (pull-out test). In every case, the samples had pull-out strengths that easily exceeded the most stringent I

requirement of Regulatory Guide 1.121 (which specifies a ininimum load capability of 3 times the normal operating value).

The staff finds that the leakage and structural capability tests are consistent with previous tests of rolled joints and that they meet all ,

applicable design and regulatory requirements. '

3.2 Deletion of Seal Weld from Tubesheet Joint l The existing Farley TS section for laser welded sleeves specifies a lower rolled joint, which included a seal weld. As demonstrated through the tests discussed in WCAP-13088, a rolled joint is adequately leak limiting and, thus, use of a seal weld in conjunction with the lower rolled joint is now regarded as unnecessary.

The seal weld practice originated from the desire to preclude any leakage through the lower rolled joint. However, numerous tests performed by 1 Westinghouse have consistently demonstrated that any rolled joint leakage,

should it occur, is a minuscule fraction of the 10 CFR Part 100 limits.

Additionally, extensive operating experience with thousands of sleeves, installed only with rolled joints, has demonstrated actual performance to be essentially leak tight.

3.3 Sleeve Plugging Limits The sleeve minimum acceptable wall thickness is determined using the criteria of Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" and ASME Code Section III allowable stress values and pressure stress equations. According to Regulatory Guide 1.121 criteria, an allowance for

! nondestructive evaluation (NDE) uncertainty and postulated operational growth of tube wall degradation within the sleeve must be accounted for when using NDE to determine sleeve plugging limits. Therefore, a conservative tube wall combined allowance for postulated degradation growth and eddy current

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-7 uncertainty of 20% through-wall per cycle was assumed for the purpose of determining the sleeve plugging limit. The sleeve plugging limit, which was calculated based on the most limiting of noma 1, upset, or faulted conditions i for welded sleeves installed at the Farley plant, was determined to be 24% of 3

i the sleeve nominal wall thickness based on ASME Code minimum material properties in accordance with staff positions. This plugging limit of 24%

provides assurance that pressure boundary integrity will be maintained.

i 3.4 Technical Specification Changes The staff finds acceptable the following proposed technical and editorial i changes to the plant TS SR 4.4.6.4 and associated bases.

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1. Table 4.4-3, " Steam Generator Repair Tube Inspection" has been i incorporated. This TS change increases the sample size requirements i from 3% to 20%, which will ensure adequate detection of additional tube i

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degradation. In addition, TS SRs have been modified to incorporate a reference to Table 4.4-3.

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2. The definition of tube repair has been modified to indicate that laser

' welded sleeves will be installed as described in Westinghouse reports WCAP-13088, Revision 4 and WCAP-14740 (January 1997).

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3. The definition of plugging or repair limit and the associated _ bases has

.i been modified to incorporate the plugging limit of 24%. This change provides assurance that pressure boundary integrity will be maintained.

, 4. The reference to a one-cycle implementation of L*, which expired at the i last Unit 2 outage has been deleted from Unit 2 TS 4.4.6.4 and 2

associated bases. In addition, the definition of L* and associated j bases have been deleted from the Unit 2 TSs.

The staff finds these TS changes consistent with the laser welded sleeve installation.

4.0 STAFF CONCLUSIONS i The staff concludes the proposed W 1aser welded sleeves, as described in the i sleeve topical reports WCAP-13088, Revision 4, and WCAP-14740, will provide sleeved tubes of acceptable metallurgical properties, structural integrity, leak tightness, and corrosion resistance. The staff also finds the preservice

and inservice inspection methods, minimum percentages, and expansion criteria

for examining the welds and sleeved tubes are acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of Alabama official l was notified of the proposed issuance of the amendments. The State official j had no comments.

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6.0 ENVIRONNENTAL CONSIDERATION The amendments change a requirement with respect to installation or use of a

facility component located within the restricted area as defined in 10 CFR Part 20 and change the surveillance requirements. The NRC staff has j determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released i offsite, and that there is no significant increase in individual or cumulative )

occupational radiation exposure. The Connission has previously issued a

proposed finding that the amendments involve no significant hazards l

consideration, and there has been no public comment on such finding l l (January 29, 1997, 62 FR 4355). The amendments also revise recordkeeping and j

, reporting requirements. Accordingly, the amendments meet the eligibility 1 1' criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (10). l Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental

! assessment need be prepared in connection with the issuance of the amendments.

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7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, i and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: G. Hornseth )

Date: March 24,1997 l

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