ML13331B020
| ML13331B020 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 03/10/1988 |
| From: | Baskin K Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13331B019 | List: |
| References | |
| TAC-67463 NUDOCS 8803160306 | |
| Download: ML13331B020 (15) | |
Text
- Enclosure 1 BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON
)
COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY )
for a Class 104(b) License to Acquire,
)
DOCKET NO. 50-206 Possess, and Use a Utilization Facility as
)
Part of Unit No. 1 of the San Onofre Nuclear )
Amendment No. 149 Generating Station
)
SOUTHERN CALIFORNIA EDISON COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY, pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 149.
This amendment consists of Proposed Change No. 182 to Provisional Operating License No. DPR-13. Proposed Change No. 182 modifies the Technical Specifications incorporated in Provisional Operating License No. DPR-13 as Appendix A.
Proposed Change No. 182 is a request to revise Appendix A Technical Specification Section 4.16, "Inservice Inspection of Steam Generator Tubing" to include a modified plugging limit for steam generator tubes with imperfections in a region of the tube that do not affect plant safety.
The adoption of this modified limit will allow tubes to remain in service that would needlessly be plugged under the existing criteria. The use of this modified plugging limit for San Onofre Unit 1 is consistent with other Westinghouse plants that have roll transition cracking problems.
In the event of conflict, the information in Amendment Application No. 149 supersedes the information previously submitted.
8803160306 880310 PDR ADOCK 05000206 PDR
-2 Based on the significant hazards analysis provided in the Description of Proposed Change and Significant Hazards Analysis of Proposed Change No. 182, it is concluded that (1) the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
Pursuant to 10 CFR 170.12, the fee of $150 is herewith remitted.
-3 Subscribed on this _
day of /,wg.4 1988.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By:
- Knneth P. Baskin Vice President Subscribed and sworn to before me this day of 2?44.,d /'ff Notoy Public in and for the County of Lo Angeles, State of California My Commission Expires: 2d i/z/77 OFFICIAL SEAL AGNES CRABTREE Notary Public-California Charles R. Kocher UOS ANGELES COUNTY My Comm. Exp. Sep. 14, 1990JaeA.Boto LOSANELESCUN~yJames A. Beoletto MO."9 Attorneys for Southern California Edison Company By:
0 g
-4 Subscribed on this day of
, 1988.
Respectfully submitted, SAN DIEGO GAS & ELECTRIC COMPANY B*
Gary tton Senior Vice President Subscribed and sworn to before me this day of Notary Publ c in and for the County of San Diego, State of California My Commission Expires:
d_?_0_/9_
David R. Pigott Samuel B. Casey Orrick, Herrington & Sutcliffe Attorneys for San Diego Gas & Electric Company By:
David R. Pigott
9 0
I
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN
)
CALIFORNIA EDISON COMPANY
)
and SAN DIEGO GAS & ELECTRIC
)
Docket No. 50-206 COMPANY (San Onofre Nuclear
)
Generating Station Unit No. 1 CERTIFICATE OF SERVICE I hereby certify that a copy of Amendment Application No. 149 was served on the following by deposit in the United States Mail, postage prepaid, on the 10th day of March, 1988.
Henry J. McGurren, Esq.
Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C. 20545 David R. Pigott, Esq.
Samuel B. Casey, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P.O. Box 60860, Terminal Annex Los Angeles, California 90060 Michael L. Mellor, Esq.
Thelen, Marrin, Johnson & Bridges Two Embarcadero Center San Francisco, California 94111 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102
-2 C. 3. Craig' Manager U. S. Nuclear Projects I ESSD Westinghouse Electric Corporation Post Office Box 355 Pittsburgh, Pennsylvania 15230 A. I. Gaede 23222 Cheswald Drive Laguna Niguel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Jame 'A. Beolet o
DESCRIPTION AND SIGNIFICANT HAZARDS ANALYSIS OF PROPOSED CHANGE NO. 182 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-13 This is a request to revise Section 4.16, "INSERVICE INSPECTION OF STEAM GENERATOR TUBING," Appendix A Technical Specifications for San Onofre Nuclear Generating Station Unit 1 (SONGS 1).
DESCRIPTION OF CHANGE Technical Specification 4.16 requires periodic inspections of the steam generator tubing to assure its integrity for continued plant operation.
The specification delineates the scope of each inspection, frequency, corrective actions and reporting requirements. Included in the corrective actions is a specification that all tubes with imperfections that exceed the plugging limit shall be repaired or plugged.
Proposed Change No. 182 would revise Technical Specification 4.16 to include a modified tube plugging limit for tubes with imperfections in the mechanically expanded portion of the tube within the tubesheet and update the technical specification bases for this criteria.
The modified tube plugging limit would allow tubes with imperfections below an F* distance in the tube roll expansion to remain in service, provided that there are no indications of cracking in the F* distance. The revision includes definitions of the tube roll expansion, the F* distance, modification of the tube plugging limit and a requirement to report the results of inspections of all tubes in service with defects below the F* distance. Revisions are also proposed to the basis for Specification 4.16. These revisions are for consistency with the body of the specification and are considered to be administrative in nature.
SAFETY EVALUATION The following information is a review of the F* criterion implementation to determine whether or not it constitutes an unreviewed safety question as defined in 10 CFR 50.59. This determination is not required for an amendment application, but is offered as supporting information.
This safety evaluation applies to the San Onofre Nuclear Generating Station Unit 1 (SONGS 1), Westinghouse Series 27 steam generators and assesses the integrity of the tube bundle for tube eddy current inspection indications occurring in the roll expanded region of tubes within the tubesheet.
A modified tube plugging criteria has been developed for use in determining whether or not repairing or plugging of partial depth hardroll expanded tubes is necessary for degradation which has been detected in the mechanically expanded portion of the tube which is within the tubesheet. Existing SONGS 1 steam generator tube plugging and repair criteria apply throughout the tube length but do not take into account the reinforcing effect of the tubesheet on the external surface of the tube.
The Series 27 steam generators at SONGS 1 were fabricated with a partial depth roll expansion in the lower end of the tube above the tube to tubesheet weld.
Plugging criteria have been developed for tube degradation in the tube expansion region below the transition of the expanded to unexpanded portions of the tube. The elastic preload and friction between the tube and the tubesheet due to the mechanical hardroll provides the force required to resist
-2 pullout. The presence of the tubesheet acts to constrain the tube and complement its integrity in that region by essentially precluding tube deformation beyond its expanded outside diameter, i.e., the resistance to both tube rupture and collapse is significantly strengthened by the tubesheet. In addition, the proximity of the tubesheet significantly affects the leak behavior of through wall tube crack in this region.
The primary criterion included in the modified tube plugging limit developed for roll expanded tubes is designated the F* criterion. The F* criterion represents a length, designated the F* distance, of continuous roll expansion in the tubesheet such that tube pullout would not occur during either normal operation or postulated accident condition loadings.
The F* distance is measured from the bottom of the transition between the roll expansion and the unexpanded tube.
Evaluation As neither tube burst or collapse can occur within the tubesheet region, steam generator tube integrity has been assessed on the basis of both tube pullout and primary to secondary leakage considerations.
In order to evaluate the F* criterion concept for indications within the tubesheet, it was postulated that a circumferential severance of a tube could occur, which is contrary to existing plant experience. Based on plant operation and laboratory experience, the configuration of any cracks, should they occur, is axial.
For axial or nearly axial indications in the tubesheet region, the tube end remains structurally intact minimizing any potential for tube pullout. Implicit in assuming a circumferential severance to occur in the development of the F* criterion is the consideration that degradation of any extent or orientation within the tubesheet below the F* distance is demonstrated to be acceptable during normal and postulated accident conditions.
An evaluation consisting of analysis and testing programs aimed at qualifying and verifying the residual radial preload of Westinghouse steam generator tubes hardrolled into the tubesheet has been conducted to determine the length of hardroll engagement necessary to resist tube pullout forces during normal and faulted condition loadings. The required engagement length, F* distance, of roll expansion to preclude tube pullout under normal operation loading conditions was calculated, with a safety factor of 3, to be 0.75 inches.
This F* distance is measured from the bottom of the transition between the roll expansion and the unexpanded portion of the tube and does not include an allowance for eddy current measurement uncertainty. The required engagement length of roll expansion to preclude tube pullout under postulated accident conditions was also calculated with appropriate safety factors and determined to be less than the value calculated for normal operation.
The stress levels in the tube above the elevation located the distance F*
below the bottom of the roll transition in the event of a circumferential break below F* would not be any larger than for a tube without a break. Thus the stress levels would not be in excess of allowable limits for normal and postulated accident conditions. Based on the determination of required
-3 engagement length and acceptable stress levels, the probability of an accident previously evaluated is not increased by the use of the alternate plugging criteria. Use of the modified tube plugging limit could not result in the consequences of a hypothetical tube rupture accident more severe than the analyzed case of a double ended guillotine break of the tube.
The engagement length determination method was derived from preload testing and was verified as conservative by both tube pullout and hydraulic proof (pressure) testing. Specifically, the F* criterion was calculated from a derived preload force and a conservative static coefficient of friction for tube to tubesheet contact. Both the tube pullout and hydraulic proof testing conducted on rolled joints provided support for the derived preload force.
Also, in assessing the F* criterion, it has been verified that the radial preload resulting from the roll is sufficient to significantly restrict leakage during normal operating and postulated accident condition loadings.
Use of the F* criterion does not change the operating conditions or function of the steam generator or any other component in the plant. The tube rupture accident bounds any condition related to tube integrity or degradation.
Therefore, the use of the modified tube plugging limit does not provide a mechanism for any accident not previously analyzed.
The factors of safety used in the development of the F* distance outlined above are consistent with the safety factors found in the ASME Boiler and Pressure Vessel Code and NRC Regulatory Guide 1.121.
Thus, the use of the F*
criterion does not involve a reduction in a margin of safety.
Conclusions On the basis of this evaluation, it is determined that tubes with any tube degradation within the tubesheet region below the F* pullout criterion (0.75 inches) can be left in service. The specified distance does not include an allowance for eddy current uncertainty in locating the elevation of an indication. Tubes with tube degradation which is located a distance of less than F* below the bottom of the transition between the roll expansion and the unexpanded portion of the tube should be removed from service by plugging or repaired in accordance with plant technical specification requirements.
The criteria defined in this evaluation have been demonstrated to result in tube integrity considerations commensurate with Reg. Guide 1.121 criteria both analytically (tube preload) and empirically (tube pullout and proof testing and leakage testing). As outlined above, use of F* criterion will not increase the probability or consequences of a previously analyzed accident or result in a previously unanalyzed accident. The margin of safety which, in part, is provided by the safety factors included in the ASME Code and Regulatory Guide 1.121 is not reduced. Therefore, implementation of the F*
criterion within the steam generators of San Onofre Nuclear Generating Station Unit 1 is determined not to represent an unreviewed safety question as defined in 10 CFR 50.59(a)(2).
-4 EXISTING TECHNICAL SPECIFICATION See Attachment 1 PROPOSED TECHNICAL SPECIFICATION See Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS As required by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that a proposed license amendment to implement a modified tube plugging limit for the San Onofre Nuclear Generating Station Unit 1 (SONGS 1) steam generators represents a no significant hazards consideration. In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amendment was analyzed using the following standards and found not to:
- 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.
This change is proposed due to eddy current indications of tube degradation in the mechanical roll expanded portion of the tubes within the tubesheet in the steam generators at SONGS 1. These steam generators were fabricated with a partial depth mechanical roll at the bottom of the tube. It has been determined through interpretation of eddy current examinations that the tube degradation occurring in the SONGS 1 steam generators is of the type associated with primary water stress corrosion cracking (PWSCC). It can be shown that tube plugging or repair is not required in many cases to maintain tube bundle integrity. Using the existing Technical Specification tube plugging limit, many of the tubes with these indications would have to be removed from service. The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging operations. The proposed amendment would minimize the loss of margin in the reactor coolant flow through the steam generator in LOCA analyses. The proposed amendment would avoid loss of margin in reactor coolant system flow and therefore assist in assuring that minimum flow rates are maintained in excess of that required for operation at full power.
Reduction in the amount of tube plugging required can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.
The possibility of tube repair by sleeving should not be considered a reason to exclude use of the F* criterion, but should be considered one of the options used to address degradation in the expanded region of the tube. The disadvantages of tube plugs noted above also apply to some extent to sleeves.
Additionally, installation of sleeves involves some impact on eddy current testing due to the changes in geometry at the ends and expansions of the sleeve and the size of probe that can pass through the reduced diameter of the sleeve.
-5 The proposed change addresses the action required when degradation has been detected in the mechanically expanded hardroll portion of steam generator tubes within the steam generator tubesheet. Existing tube repair or plugging criteria, i.e., current applications of USNRC Regulatory Guide 1.121, do not take into account the effect of the tubesheet on the external surface of the tube. The presence of the tubesheet will enhance the integrity of degraded tubes in that region by precluding tube deformation beyond the expanded outside diameter. Additionally, a portion of the expansion at the top end of the roll expansion is sufficient to preclude pullout of the tube during normal operation and postulated accident condition loadings if a tube were postulated to sever circumferentially during plant operations in the portion of the tube covered by the proposed amendment. Finally, the roll expansion of the tube into the tubesheet provides a barrier to significant leakage for through wall cracking of the tube in the expanded region.
The proposed change designates a portion of the tube for which tube degradation does not necessitate remedial action, except as dictated for compliance with tube leakage limits as set forth in the SONGS 1 Technical Specifications. As noted above, the area subject to this change is in the mechanically expanded portion of the tube within the tubesheet of the steam generators. The length of mechanical expansion required to resist pullout for all postulated conditions, designated the F* distance, has been determined to be 0.75 inches.
For the purposes of the evaluation of the F* distance, the unexpanded portion of the tube was assumed not to provide resistance to pullout or leakage. Since the roll expansion of the tube above the F*
distance is sufficient to preclude pullout of the tube, use of the F* criteria does not depend on any determination of the condition of tube degradation in the portion of the tube below the F* distance.
The proposed amendment would modify Technical Specifications 4.16, "INSERVICE INSPECTION OF STEAM GENERATOR TUBING," which provides tube inspection requirements and acceptance criteria to determine the level of degradation for which the tube may remain in service. The proposed amendment would add definitions required for the modified tube plugging limit, prescribe the portion of the tube subject to the criteria and require reporting of the results of inspections of those tubes in service, with defects below the F*
distance. The proposed Technical Specification defines the F* distance to be 1.0 inches versus the 0.75 inches justified above, as an eddy current measurement error of 0.25 inches is included in the proposed specifications.
The proposed Technical Specification changes accompany this analysis.
Analysis Conformance of the proposed amendments to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:
- 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
-6 RESPONSE: NO The supporting technical and safety evaluations of the subject criteria demonstrate that the presence of the tubesheet will enhance the tube integrity in the region of the hardroll by precluding tube deformation beyond its initial expanded outside diameter. The resistance to both tube rupture and tube collapse is strengthened by the presence of the tubesheet in that region. The result of the hardroll of the tube into the tubesheet is an interference fit between the tube and the tubesheet. Tube rupture cannot occur because the contact between the tube and tubesheet does not permit sufficient movement of tube material. In a similar manner, the tubesheet does not permit sufficient movement of tube material to permit collapse of the tube during postulated LOCA loadings.
Additionally, through analysis and testing, it has been demonstrated that the roll expansion above the F* distance is sufficient to preclude pullout of the tube from the tubesheet. Even with the conservative assumption that a tube could completely sever circumferentially below the F* distance, test results demonstrate that pullout of the tube is precluded under normal and postulated accident condition loadings.
This assumption is conservative as the PWSCC that has been observed in operating units has been typified as short and axially oriented.
Relative to expected leakage, the length of roll expansion above the F* distance is sufficient to preclude significant leakage from tube degradation located below the F* distance. The existing Technical Specification leakage rate requirements and accident analysis assumptions remain unchanged in the unlikely event significant leakage from this region does occur. As noted above, tube rupture and pullout is not expected for tubes using the modified tube plugging limit. Any leakage out of the tube from within the tubesheet at any elevation in the tubesheet is fully bounded by the existing steam generator tube rupture analysis for SONGS 1. The proposed modified tube plugging limit does not adversely impact any other previously evaluated design basis accident. Therefore, it is concluded that operation of the facility in accordance with this proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
-7 RESPONSE: NO Implementation of the proposed modified tube plugging limit does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism to result in an accident outside of the region of the tubesheet expansion. Any hypothetical accident as a result of any tube degradation in the expanded portion of the tube would be bounded by the existing tube rupture accident analysis. Therefore, it is concluded that operation of the facility in accordance with this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
RESPONSE: NO The use of the modified tube plugging limit has been demonstrated to maintain the integrity of the tube bundle commensurate with the requirements of NRC Regulatory Guide 1.121 for indications in the free span of tubes and the primary to secondary pressure boundary under normal and postulated accident conditions. Acceptable tube degradation is any degradation in the tubesheet more than the F*
distance below the bottom of the transition between the roll expansion and unexpanded portion of the tube. The safety factors used in the determination of the F* distance are consistent with the safety factors in the ASME Boiler and Pressure Vessel Code used in steam generator design. The F* distance has been verified by testing to be greater than the length of roll expansion required to preclude significant leakage during normal and postulated accident conditions.
Implementation of the modified tube plugging limit will decrease the number of tubes which must be taken out of service with tube plugs or repaired with sleeves. Both plugs and sleeves reduce the RCS flow margin, thus implementation of the alternate plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of increased plugging or sleeving. Based on the above, it is concluded that the proposed change does not result in a significant reduction in a margin with respect to plant safety or the bases of the plant technical specifications. Therefore, it is concluded that operation of the facility in accordance with this proposed change does not involve a significant reduction in a margin of safety.
-8 SAFETY AND SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the preceding analysis, it is concluded that operation of San Onofre Nuclear Generating Station, Unit 1 in accordance with Proposed Change No. 182, does not involve a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change. - Existing Specifications - Proposed Specifications LAB:9360F