ML20151M128

Safety Evaluation Supporting Amends 111 & 100 to Licenses DPR-39 & DPR-48,respectively

ML20151M128
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	04/06/1988
From:	Office of Nuclear Reactor Regulation
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ML20151M120	List: ML20151M115 ML20151M119 ML20151M128 ML20196F977
References
NUDOCS 8804220248
Download: ML20151M128 (5)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.111 TO FACILITY OPERATING LICENSE NO. DPR-39 AND j AMENDHENT NO.100 TO FACILITY OPERATING LICENSE NO. DPR-48 COMMONWEALTH EDISON COMPANY ZION NUCLEAR POWER STATION. UNITS 1 AND 2 DOCKET NOS. 50-295 AND 50-304 l

1.0 INTRODUCTION

By letters dated December 24, 1987 and February 11, 1988, the Consnonwealth Edison Company, the licensee, submitted a request to change the technical specifications to allow for the installation of steam generator tube repair sleeves in the Zion Nuclear Power Station Units 1 and 2.

2.0 BACKGROUND

Zion Units 1 and 2 use pressurized water reactor (PWR) steam generators that are manufactured by the Westinghouse Electric Corporation. B0th units have four Westinghouse Series 51 vertical recirculating U-bend tube steam generators, and each has 3388 Inconel 600 tubes. The tubes have a 0.050 inch nominal wall thickness with an outside diameter of 0.875 inches. Zion 1 initiated commercial service in December 1973 and Zion 2 in September 1974.

Both Zion units have experienced steam generator tube degradation in the tube-sheet area. The degradation is believed to be corrosion induced intergranular attack (IGA). Because Zion Unit I has experienced one of the highest rates of tube plugging in Westinghouse Model 51 steam generators, Conrnonwealth Edison Company submitted a proposed Technical Specification change to the NRC on l April 24, 1986 that would allow Comt,ustion Engineering welded sleeves to be installed as an alternative to plugging for steam generator repair. On November 18, 1986, the NRC approved Commonwealth Edison Company's request to  !

perform tube sleeving according to the Combustion Engineering procedure.

3.0 DISCUSSION Westinghouse Report WCAP-11669, "Zion Units 1 and 2 Steam Generator Sleeving Report," was submitted in support of the request for approval of sleeving.

Westinghouse provides a leak limiting sleeve which is secured to the steam gene-rator tube near each end of the sleeve. The sleeve spans the degraded area of the parent steam generator tube in and above the tubesheet region.

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2 The operation of pressurized water reactor steam generators het, in some in-stances, resulted in localized corrosive attack on the inside (primary side) or outside (secondary side) of the steam generator tubing. This corrosive attack results in a reduction in steam generator tube wall thickness. Steam generator tubing has been designed with considerable margin between the actual wall thick-ness anc the wall thickness required to meet structural requirements. Thus, it has not been necessary to take corrective action unless structural limits are  :

being approached.

Historically, the corrective action taken where steam generator tube wall de-gradation has been severe, has been to install plugs at the inlet and outlet ends of the steam generator tube when the reduction in wall thickness reaches a cal-culated value referred to as the plugging limit. Eddycurrenttesting(ECT)has been used to measure steam generator tubing degradation and the tube plugging limit takes into account ECT measurement uncertainty.

1 The objectives of installing sleeves in the steam generator tubes are twofold.

The sleeve must maintain the structural integrity of the steam generator tube during normal operating and postulated accident conditions. Additionally, the sleeve must prevent leakage in the event of a through hole in the wall of the steam generator tube. Tests and analyses were perfonned to demonstrate the capability of the sleeves to perform these functions under nonnal and postulated accident conditions. Plugs are installed in the steam g2nerator tubes when the tubes cannot be successfully repaired with sleeves. l The sleeve designed by Westinghouse is inserted in a degraded or defective tube j 4 and spans the degraded region of the original tube. The sleeve length is con-trolled by the insertion clearance between the channel head inside surface and 1 the primary side of the tubesheet, and the tube degradation location above the l tubesheet. The remaining design parameters such as wall thickness and material i are selected to enhance design margins and corrosion resistance and to meet ASME Boiler and Pressure Vessel Code requirements. The upper joint is located so as to provide a f ree length of sleeve above the degraded region of the original tube. This length is added so that if in the unlikely event the existing tube were to become severed just above the upper edge of the joint, the tube would be restrained by the sleeve from lateral and axial motion, and subsequent leakage i would also be limited.

The Zion steam generators were built to the 1965 Edition of Section III of the ASME Boiler and Pressure Vessel Code; however, the sleeves have been designed and analyzed b6 sed upon the 1983 Edition of Section III of the Code through the Winter 1983 Addenda as well as applicable Regulatory Guides. The associated materials and processes also meet the Code requirements and criteria.

The sleeve material, thermally treated (TT) Inconel 690, was selected to provide additional resistance to stress corrosion cracking. Evaluations were performed by Westinghouse which (1) verified that thermally treated Inconel 690 is a suit-able material for use in steam generator environments, and (2) verified that sleeving does not have a detrimental effect on the serviceability of the existing tube or the sleeve components, i

3 The sleeve installation consists of a series of steps including tube preparation, sleeve insertion, joint fonnation and finally joint inspection. In preparation for sleeve insertion, the inside diameter areas of the tubes to be sleeved are cleaned to prepare the surface for joint formaticn and removal of loose oxide and foreign material.

4.0 EVALUATION The staff evaluation of the amendment proposed by Comonwealth Edison Company and the WCAP-11669 report prepared by Westinghouse is based upon the following considerations: Steam generator tube sleeving is a repair technique that is an alternative to removing defective or degrading tubes from service by plugging. -

Sleeves are designed to span a defective or degraded region of a steem generator .

tube and to maintain the steam generator tubing primary-to-secondary pressure i boundary under normal and accident conditions. A successful sleeving system must provide a corrosion resistant sleeve material with structural integrity and leak tightness of the sleeved tube.

The staff concurs that the use of Inconel 690 TT sleeves is an improvement over the Inconel 600 material used in the original steam generator tubin Corrosion t testsconductedundertheElectricPowerResearchInstitute(EPRI)g. sponsorship l confirm WCAP-11669 test results regarding the improved corrosion resistance of Inconel 690 TT over that of Inconel 600. Accelerated stress corrosion tests in '

caustic and chloride aqueous solutions have also indicated that Inconel 690 TT resists general corrosion in aggressive environments. Isothermal tests in high purity water have shown that, at normal stress levels, Inconel 690 TT has high resistance to intergranular stress corrosion cracking in extended high tem- .

perature exposure. EPRI concluded as a result of these laboratory r.orrosion  :

tests, that Inconel 690 TT material could be used for PWR steam generator tubing with all volatile treatment secondary water systems. Inconel 690 is a Code ap-  :

proved material (ASME $8-163), covered by ASME Code Case N-20, and is acceptable to NRC under Regulatory Guide 1.85 (Rev. 24, July 1986). The NRC staff has approved Westinghouse's use of Inconel 690 TT tubing in replacement steam gene-rators to be placed in service.

Corrosion test results presented in WCAP-11669 confirm that the expansion pro-cesses used to form the upper and lower sleeve-to-tube joints, while mechani-cally working the sleeve and tube, did not impose sufficiently high residual tensile stresses as to promote stress corrosion cracking in the tube-sleeve samples used in the tests.

There was some question concerning the adequacy of the combined allowance for eddy current uncertainty and operational degradation to be used in connection with the proposed amendment. After discussions with the staff, the licensee proposed to clarify the conservatism to be used and to adopt the staffs recom-mendation of a combined allowance of 20 percent. The licensee's letter of February 11, 1988 modified the sleeve plugging limit to 31 percent wall thick-ness in order to properly reflect this conservatism. Therefore the operational sleeve thickness of 69 percent includes the minimum calculated wall thickness based on normal loading conditions, which are controlling, plus the margin for eddy current uncertainty and postulatad operational degradation in accordance with Regulatory Guide 1.121 and staff positions.

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Sleeving introduces special inspection problems in the eddy current testing of the sleeve-tube combination. First, a method confirming that the joints meet critical process dimensions is required. Second, it must be shown that the tube-sleeve assembly is capable of being evaluated through subsequent routine inservice inspection.

The staff finds that the use of the eddy current equipment and techniques as described in WCAP-11669, or their equivalent, is currently acceptable. However, the licensee has comitted to utilize advanced state of the art techniques as they are developed and verified.

The process of sleeving restores structural integrity of the degraded portions of the steam generator tubes. For that reason the steam generator's response to transients and accident scenarios previously analysed remains unchanged.

The sleeving process also does not create conditions leading to new accidents i' not previously analysed.

The Emergency Core Cooling System (ECCS) perfonnance analysis for Zion 1 and 2 supports operation with up to 10 percent of the steam generator tubes plugged out of service or the hydraulic equivalent ratio for sleeved tubes.

Implementation of the sleeving process will decrease the number of tubes which must be taken out of service with tube plugs. Both plugs and sleeves reduce the reactor cooling flow margin, but the use of sleeves will maintain a margin of flow that would othenvise be reduced in the event of plugging those tubes that are sleeved. Using the calculated sleeve to plug ratio, the Licensee has shown that for the level of sleeving projected to occur at Zion 1 and 2 no ECCS results are more adverse than those in the existing Safety Analysis. In addition, the non-LOCA analysis and the transient evaluation are not affected by the sleeving. The staff finds, therefore, that the operation of the facility in accordance with the approved sleeving process would (1) minimize the loss of margin in the reactor coolant flow through the steam generators in the LOCA analysis and (2) assure that  :

minimum flow rates are maintained in excess of that required for operation at full power. Based on the above, the staff concludes that the proposed technical specification change does not result in a significant reduction of the margin of safety.

The staff's evaluation is based in part on the enclosed Technical Evaluation Report prepared by Science Applications International Corporation.

5.0 TECHNICAL FINDING The staff conclusions are based on a review and evaluation of the infonnation ,

and data presented in WCAP-11669, "Zion Units 1 and 2 - Steam Generator Sleeving I Report (Mechanical Sleeves)" dated December 1987. The staff concludes that the i request by Commonwealth Edison Company for a proposed amendment to Facility 1 License Nos. DPR-39 and DPR-48, Appendix A. Section 43, "Reactor Coolant System,"

that will allow Zion's steam generator tubes to be repaired by utilizing West-inghouse Electric Corporation mechanical sleeving methodology is acceptable.

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6.0 ENVIRONMENTAL CONSIDERATION

These amendments involve a change in the installation or use of the facilities component located with the restricted areas as defined in 10 CFR 20. The staff has determined that these amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released of f site, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public connent on such finding. Accord-ingly, these amendments meet the eligibility criteria for categorical exclusion setforthin10CFR51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.

7.0 CONCLUSION

The staff 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, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: H. Conrad Dated: April 6, 1988

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