ML12222A340
| ML12222A340 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 07/31/2012 |
| From: | Price J Dominion, Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 12-487, FOIA/PA-2013-0030, FOIA/PA-2013-0139 | |
| Download: ML12222A340 (35) | |
Text
10 CFR 50.90 VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 July 31, 2012 Attention: Document Control Desk Serial No.12-487 U. S. Nuclear Regulatory Commission NL&OS/ETS RO 11555 Rockville Pike Docket Nos.
50-280/281 Rockville, MD 20852 License Nos.
DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
SURRY POWER STATION UNITS I AND 2 PROPOSED TECHNICAL SPECIFICATIONS TO ADOPT TSTF-510, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Pursuant to 10 CFR 50.90, Dominion requests an amendment in the form of changes to the Technical Specifications (TS) to Facility Operating Licenses DPR-32 and 37 for Surry Power Station (SPS) Units 1 and 2, respectively.
The proposed amendment would modify the TS requirements regarding steam generator tube inspections and reporting as described in TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
The availability of this Traveler was announced in the Federal Register on October 27, 2011 (72 FR 66763) as part of the consolidated line item improvement process (CLIIP).
Because SPS has not adopted Standard Technical Specifications (STS), Dominion is proposing minor variations and/or deviations from the TS changes described in TSTF-510, Revision 2, to provide consistent terminology and format with the SPS TSs. The minor variations and/or deviations from the specific wording/format provided in TSTF-510, Revision 2, are considered administrative and do not change the meaning, intent, or applicability of the CLIIP. provides a description and assessment of the proposed changes including: the requested plant-specific licensing basis that is equivalent to the 10 CFR 50, Appendix A General Design Criteria referenced in the Traveler, as well as the plant-specific administrative variations from the TS changes described in TSTF-510, Revision 2. Attachment 2 provides the current TS pages and Bases pages marked up to show the proposed changes. Attachment 3 provides the proposed TS pages. The TS Bases changes are for information only.
Serial No.12-487 Docket Nos. 50-280/281 SG Program Revision Page 2 of 3 The proposed amendment does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92. The Facility Safety Review Committee has reviewed and concurred with the determinations herein.
Approval of the proposed amendment is requested by July 2013. Once approved, the amendment shall be implemented within 60 days.
In accordance with 10 CFR 50.91 (b), a copy of this license amendment request is being provided to the State of Virginia.
If you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.
Very truly yours, VICKI L. HULL J.a rc Notary public J.
an Price I
Commonwealth of Virginia Vice President - Nuclear Engineering 1
140542
[
1 My Commission Expires May 31, 2014 COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. Alan Price, who is Vice President - Nuclear Engineering of Virginia Electric and,Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this day of 1L
,2012.
My Commission Expires:.
hii..1 ;0&.
Notary Public Commitments made in this letter: None
Enclosures:
- 1. Discussion of Change
- 2. Marked-up Technical Specifications and Bases Pages
- 3. Proposed Technical Specifications and Bases Pages
Serial No.12-487 Docket Nos. 50-280/281 SG Program Revision Page 3 of 3 cc:
U. S. Nuclear Regulatory Commission Region II Marquis One Tower - Suite 1200 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street Suite 730 Richmond, Virginia 23219 NRC Senior Resident Inspector Surry Power Station Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North - Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North - Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738
Serial No.12-487 Docket Nos. 50-280/281 SG Program Revision Discussion of Change Virginia Electric and Power Company (Dominion)
Surry Power Station Units I and 2
Serial No 12-487 SG Program Revision Page 1 of 5 DESCRIPTION AND ASSESSMENT
1.0 DESCRIPTION
The proposed change revises Specifications 6.4.Q, "Steam Generator (SG) Program" and 6.6.A.3, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods and to address other administrative changes and clarifications.
For consistency, additional administrative changes are being made to Specifications 3.1.H "Steam Generator (SG)
Tube Integrity," and 4.19, "Steam Generator (SG) Tube Integrity."
The proposed amendment is consistent with TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
2.0 ASSESSMENT
2.1 Applicability of Published Safety Evaluation Dominion has reviewed TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"
(ADAMS Accession No. ML110610350) and the model safety evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513) as identified in the Federal Register Notice of Availability, dated October 27, 2011 (76 FR 66763). As described in the subsequent paragraphs, Dominion has concluded that the justifications presented in TSTF-510 and the model safety evaluation prepared by the NRC staff are applicable to Surry Power Station (SPS) Units 1 and 2 and justify this amendment for the incorporation of the changes to the SPS Unit 1 and 2 Technical Specifications (TS).
2.2 Optional ChanQes and Variations Dominion is not proposing any technical variations or deviations from the TS changes described in the TSTF-51 0, Revision 2, or the applicable parts of the NRC staff's model safety evaluation dated October 27, 2011.
However, Dominion is proposing the following administrative variations from the TS changes described in the TSTF-510, Revision 1, or the applicable parts of the NRC staff's model safety evaluation dated October 27, 2011.
The SPS custom TS numbering system is different than the Improved Technical Specifications (ITS) on which TSTF-510 was based. Specifically, the "SG Program" in the Surry TS is numbered 6.4.Q rather than 5.5.9; the "Steam Generator (SG) Tube Integrity" TS is numbered 3.1.H rather than 3.4.17; and the "Steam Generator Tube Inspection Report is numbered 6.6.A.3 rather than 5.6.7.
In addition the Steam Generator (SG) Tube Integrity surveillances are located in TS 4.19. These differences are administrative and do not affect the applicability of TSTF-51 0 to the Surry TS.
Serial No 12-487 SG Program Revision Page 2 of 5 The proposed change also corrects an administrative inconsistency in TSTF-510, Paragraph d.2 of the Steam Generator Tube Inspection Program.
In Section 2.0, "Proposed Change," TSTF-510 states that references to "tube repair criteria" in Paragraph d.2 is revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted:
"If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated" (emphasis added).
The underlined phrase should state "tube plugging [or repair] criteria," consistent with the other changes made in TSTF-510.
Dominion is changing the phrase to "tube plugging criteria." This change is administrative and should not result in this application being removed from the Consolidated Line Item Improvement Process.
This administrative error was indentified in a February NRC-TSTF meeting and documented in a letter from the TSTF to the NRC dated March 28, 2012 (TSTF letter No. 12-09).
3.0 REGULATORY ANALYSIS
3.1 No Siqnificant Hazards Consideration Determination Dominion requests adoption of an approved change to the standard technical specifications (STS) into the plant specific technical specifications (TS), to revise Specifications 6.4.Q, "Steam Generator (SG) Program," 6.6.A.3, "Steam Generator Tube Inspection Report," and 4.19, "Steam Generator (SG) Tube Integrity," to address inspection periods and other administrative changes and clarifications.
As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis.
The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are
Serial No 12-487 SG Program Revision Page 3 of 5 inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of SGTR.
Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation.
The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.
Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.
Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.
Serial No 12-487 SG Program Revision Page 4 of 5 Based on the above, Dominion concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
3.2 Applicable Regulatory Requirements/Criteria During the initial plant licensing of SPS Units 1 and 2, it was demonstrated that the design of the reactor coolant pressure boundary met the regulatory requirements in place at that time. The General Design Criteria (GDC) included in Appendix A to 10 CFR Part 50 did not become effective until May 21, 1971. The Construction Permits for SPS Units 1 and 2 were issued prior to May 21, 1971; consequently, these units were not subject to GDC requirements.
(Reference SECY-92-223 dated September 18, 1992.) However, the following information demonstrates compliance with GDC 14, 15, 30, 31, and 32 of 10 CFR 50, Appendix A. Specifically, Section 1.4 of the UFSAR discusses the design of the station relative to the design criteria published in 1971.
The GDC state that the Reactor Coolant Pressure Boundary (RCPB) shall have "an extremely low probability of abnormal leakage
. and gross rupture" (GDC 14/UFSAR 1.4.9),
"shall be designed with sufficient margin" (GDCs 15/UFSAR 1.4.33 and 31/UFSAR 1.4.34), shall be of "the highest quality standards practical" (GDC 30/UFSAR 1.4.1), and shall be designed to permit "periodic inspection and.testing
. to assess
. structural and leak tight integrity" (GDC 32/UFSAR 1.4.36). Structural integrity refers to maintaining adequate margins against burst and collapse of the SG tubing. There are no changes to the SG design that impact these general design criteria.
The TS repair limits ensure that tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions. The reactor coolant pressure boundary is designed, fabricated and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime.
Reactor coolant pressure boundary components have provisions for the inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. Structural integrity refers to maintaining adequate margins against burst and collapse of the SG tubing.
Leakage integrity refers to limiting primary-to-secondary leakage to within acceptable limits during all plant conditions.
10 CFR 50, Appendix B, establishes quality assurance requirements for the design, construction, and operation of safety related components. The pertinent requirements of this appendix apply to all activities affecting the safety related functions of these components.
These requirements are described in Criteria IX, Xl, and XVI of Appendix B and include control of special processes, inspection, testing, and corrective action.
Under 10 CFR 50.65, the Maintenance Rule, licensees classify SGs as risk significant components because they are relied upon to remain functional during and after design
Serial No 12-487 SG Program Revision Page 5 of 5 basis events. SGs are to be monitored under 10 CFR 50.65(a)(2) against industry established performance, criteria.
Meeting the performance criteria of NEI 97-06, Revision 3, provides reasonable assurance that the SG tubing remains capable of fulfilling its specific safety function of maintaining the reactor coolant pressure boundary.
4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
Serial No.12-487 Docket Nos. 50-280/281 SG Program Revision Marked-up Technical Specifications and Bases Pages (Bases for Information Only)
Virginia Electric and Power Company (Dominion)
Surry Power Station Units 1 and 2
TS 3.1-26 H.
Steam Generator (SG) Tube Integrity Applicability The following specifications are applicable whenever Tavg (average RCS temperature) exceeds 200'F (200 degrees Fahrenheit).
Specifications ping
- 1.
SG tube integrity shall be maintained, and all SG tubes satisfying the tube e criteria shall be plugged in accordance with the Steam Generator Program.
- 2.
If the requirements of 3.1.H. 1 are not met for one or more SG tubes, then perform the following:'
- a. Within 7 days, verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection; and
- b. Plug the affected tube(s) in accordance with the Steam Generator Program prior to Tavg exceeding 200'F following the next refueling outage or SG tube inspection.
- 3.
If the required actions of Specification 3.1.11.2 are not completed within the specified completion time, or SG tube integrity is not maintained, the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Note:
- 1.
A separate TS action entry is allowed for each SG tube.
BASES BACKGROUND - Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.1.A.2.
SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.
Amendment Nos. 25+ nd 250
TS 3.1-28 I
-9349-9 K
Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LIMITING CONDITIONS FOR OPERATION - The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the
-riteria be plugged in accordance with the Steam Generator Program.
plugging During an SG inspection, any inspected tube that satisfies the Steam Gen r Progrý criteria is removed from service by plugging. If a tube was determined to satisfy th -e criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.4.Q, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that significantly affect burst or collapse. In that context, the term "significantly" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.
Amendment Nos. 251 rtnd2"5O
TS 3.1-30 3.1l.H.2.a andb pugn Specification 3.1.H.2 applies if it is discovere one or more SG tut es examined in an inservice inspection satisfy the tube r riteria but were not plugged in ccordance with the Steam Generator Program as required by SR 4.19. An evaluation of SG t be integrity of the affected tube(s) must be made. Steam generator tube integrity is based n meeting the SG performance criteria described in the Steam Generator Program. The SG wipa criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Specification 3.1.H.3 applies.
A completion time of 7 days is sufficient to complete the evaluation while minimizing the risk of unit operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity, required action 3.1.H.2.b allows unit operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to Tavg exceeding 200'F following the next refueling outage or SG inspection. This completion time is acceptable since operation. until the next inspection is supported by the operational assessment.
3.1.H.3 If the required actions and associated completion times of Specification 3.1.H.2 are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The allowed completion times are reasonable, based on operating experience, to reach the desired unit conditions from full power conditions in an orderly manner and without challenging plant systems.
Amendment Nos. 25-and-25 E
TS 4.19-1 4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Applicability Applies to the verification of SG tube integrity in accordance with the Steam Generator Program.
Objective To provide assurance of SG tube integrity.
Specifications plugging A.
Verify SG tube integrity in accordance with the Steam GeneratProgram.
B.
Verify that each inspected SG tube that satisfies the tube epi criteria is plugged in accordance with the Steam Generator Program prior to Tavg exceeding 200'F following a SG tube inspection.
BASES SURVEILLANCE REQUIREMENTS (SR)
SR 4.19.A During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
plugging The Steam Generator Program determines the scope of the inspe tion and the methods used to determine whether the tubes contain flaws satisfying the tube fP4,t-criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.
Amendment Nos. +-and--250
TS 4.19-2 The Steam Generator Program defines the frequency of SR 4.19.A. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 7). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.4.Q contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.
n s SR 4.19.13 plugging During an SG inspection, any inspected tube th satisfie the am Generator Program criteria is removed from service by p1 ging. The ube reei. criteria delineated in Specification 6.4.Q are intended to ensurc at tubes accept for continued service satisfy the SG performance criteria with allowan or error in the flaw s ze measurement and for future flaw growth. In addition, the tube Pepe4 criteria, in conjunctio with other elements of the Steam Generator Program, ensure that the SG performance criteria ill continue to be met until the next inspection of the subject tube(s). Reference 1 and Reference 7 provide guidance for performing operational assessments to verify that the tubes remaining in s rvice will continue to meet the SG performance criteria.
The frequency of prior to Tavg exceeding 200'F following SG inspection ensures that the Surveillance has been completed and all tubes meeting the rPOP6 criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.
REFERENCES
- 1. NEI 97-06, "Steam Generator Program Guidelines."
- 2.
- 3.
- 4.
Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
- 5.
ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.
- 6.
Draft Regulatory Guide 1.12 1, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.
- 7.
EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."
Amendment Nos. 25+1anl-250
Insert B for TS Bases - Surveillance Requirements If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.4.Q until subsequent inspections support extending the inspection interval.
TS 6.4-11 29=07 Q. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following p:
- 1. Provisions for condition monitoring assessments.
Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
- 2. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- a. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and Z
ow aod all anticipated transients included in the design specificatiori
~ýdesign basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- b. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
Leakage is not to exceed 1 gpm for all SG.
Amendment Nos. 251rand--25
TS 6.4-12
- c. The operational LEAKAGE performance criterion is specified in TS 3.1.C and 4.13, "RCS Operational LEAKAGE."
pluqqinq
- 3. Provisions for SG tube e'Sariteria. T es found by inservice inspection to contain flaws with a depth equal to exceeding 40% of the nominal tube wall thickness shall be plugged.
The following alternate tube re*r criteria shall be applied as an alternative to the 40% depth-based criteria:
- a. Tubes with service-induced flaws located greater than 17.89 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 17.89 inches below the top of the tubesheet shall be plugged upon detection.
Amendment Nos. 2R -and- ý7*
TS 6.4-13 04-O17t2-
- 4. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the ltube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube criteria. Portions of the tube greater than 17.89 inches below the top of the tubesheet are excluded from this requirement. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4.a, 4.b, and 4.c below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. Ayf of degradation shall be performed to determine the type and location of flaw s t
'4+t~
e t tubes m ay be susceptible and, based on this assessm ent, to aesen d-etermine which inspection methods need to be employed and at what locations.
- a. Inspect 100% of the tubes in each SG durn the first refueling outage fo llo w in g S G a.
m 49-in sta lltoI.atio n
- b. Ins ect 100% of th tubes at sequential perio s of 120, 90, and, hereafter, 60 effec 've full power m ths. The first sequentia eriod shall be co sidered to begin aft the first inservi inspection of the SGs. n addition, inspec 0% of the tubes b yhe refueling outme nearest the midpoiof the period anthe remaining 50%
the refueling ou e nearest the end he period. No th shall operarte for m e than 48 effecti full power month o two refueling o~uta es (w hichever is le s wt o tb i i s ec d
- c. If crack indications are found in the portions of the SG tube not excluded above, then the next inspection for eachSG for the degradation mechanism that caused the crack indication shall ot exceed 24 effective full power monhs r oe rfueingoutge(whiche er
-*). If definitive information, rmainngh 50%
the refueling out ge nersthed tepridN such as from examination of a pulled tr te, inostic non-destructive testing, or engineering evaluation indicates hat a fack-like indication is not associated with a crack(s), then the indi ation nee ot be treated as a crack.
- 5. Provisions for monitoring operational prim ry to seconda LEAKAGE.
affected and potentially affected results in more frequent inspections Amendment Nos.
and ý-
Insert A - Surry TS 6.4.Q SG Program (600TT)
- b.
After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in b.1, b.2, and b.3 below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
- 1)
After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period;
- 2)
During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and
- 3)
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.
TS 6.6-3
- b. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.1.D.4. In addition, the information itemized in Specification 3.1.D.4 shall be included in this report.
- 3. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after Tavg exceeds 200'F following completion of an inspection performed in accordance with the Specification 6.4.Q, Steam Generator (SG) Program. The report shall include:
- a. The scope of inspections performed on each SG,
- b. Aefti,*e*gradation mechanisms found,
- c. Nondestructive examination techniques utilized for each degradation mechanism,
- d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e. Number of tubes plugged during the inspection outage for each aioe--
degradation mechanism,
- f. Tctal numfibor and porcontago of twbes pluggid to dat,
- g. The results of condition monitoring, including the results of tube pulls and in-situ testing,
'i The primary to secondary LEAKAGE rate observed in each SG (if it is not 4
practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, and The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator.
Amendment Nos. 277 and 277
TS 6.6-3a I.
The calculated accident induced LEAKAGE rate from the portion of the tubes below 17.89 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced LEAKAGE rate from the most limiting accident is less than 1.80 times the' maximum operational primary to secondary LEAKAGE rate, the report should describe how it was determined/, -and I
The results of the monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
Amendment Nos. 27,and 27-
Serial No.12-487 Docket Nos' 50-280/281 SG Program Revision Proposed Technical Specification and Bases Pages (Bases for Information Only)
Virginia Electric and Power Company (Dominion)
Surry Power Station Units 1 and 2
TS 3.1-26 H.
Steam Generator (SG) Tube Integrity Applicability The following specifications are applicable whenever Tavg (average RCS temperature) exceeds 200'F (200 degrees Fahrenheit).
Specifications
- 1.
SG tube integrity shall be maintained, and all SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.
- 2.
If the requirements of 3.1.H. 1 are not met for one or more SG tubes, then perform the following:'
- a. Within 7 days, verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection; and
- b. Plug the affected tube(s) in accordance with the Steam Generator Program prior to Tavg exceeding 200'F following the next refueling outage or SG tube inspection.
- 3.
If the required actions of Specification 3.1.H.2 are not completed within the specified completion time, or SG tube integrity is not maintained, the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Note:
- 1.
A separate TS action entry is allowed for each SG tube.
BASES BACKGROUND - Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.1.A.2.
SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.
Amendment Nos.
TS 3.1-28 Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LIMITING CONDITIONS FOR OPERATION - The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.4.Q, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that significantly affect burst or collapse. In that context, the term "significantly" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.
Amendment Nos.
TS 3.1-30 3.1.H.2.a and b Specification 3.1.H.2 applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube plugging criteria but were not plugged in accordance with the Steam Generator Program as required by SR 4.19. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG plugging criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Specification 3.1.11.3 applies.
A completion time of 7 days is sufficient to complete the evaluation while minimizing the risk of unit operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity, required action 3.1.H.2.b allows unit operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to Tavg exceeding 200'F following the next refueling outage or SG inspection. This completion time is acceptable since operation until the next inspection is supported by the operational assessment.
3.1.H.3 If the required actions and associated completion times of Specification 3.1.H.2 are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The allowed completion times are reasonable, based on operating experience, to reach the desired unit conditions from full power conditions in an orderly manner and without challenging plant systems.
Amendment Nos.
TS 4.19-1 4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Applicability Applies to the verification of SG tube integrity in accordance with the Steam Generator Program.
Objective To provide assurance of SG tube integrity.
Specifications A.
Verify SG tube integrity in accordance with the Steam Generator Program.
B.
Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program prior to Tavg exceeding 200'F following a SG tube inspection.
BASES SURVEILLANCE REQUIREMENTS (SR)
SR 4.19.A During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube plugging criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.
Amendment Nos.
TS 4.19-2 The Steam Generator Program defines the frequency of SR 4.19.A. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 7). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.4.Q contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.4.Q until subsequent inspections support extending the inspection interval.
SR 4.19.B During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. The tube plugging criteria delineated in Specification 6.4.Q are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 and Reference 7 provide guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.
The frequency of prior to Tavg exceeding 200'F following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.
Amendment Nos.
TS 4.19-2a REFERENCES
- 1.
NEI 97-06, "Steam Generator Program Guidelines."
- 2.
- 3.
- 4.
Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
- 5.
ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.
- 6.
Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.
- 7.
EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."
Amendment Nos.
TS 6.4-11 Q. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:
- 1. Provisions for condition monitoring assessments.
Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
- 2. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- a. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- b. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the Ieakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
Leakage is not to exceed 1 gpm for all SG.
Amendment Nos.
TS 6.4-12
- c. The operational LEAKAGE performance criterion is specified in TS 3.1.C and 4.13, "RCS Operational LEAKAGE."
- 3. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth-based criteria:
- a. Tubes with service-induced flaws located greater than 17.89 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 17.89 inches below the top of the tubesheet shall be plugged upon detection.
- 4. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. Portions of the tube greater than 17.89 inches below the top of the tubesheet are excluded from this requirement. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4.a, 4.b, and 4.c below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.
A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- a. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
Amendment Nos.
TS 6.4-13
- b.
After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in b. 1, b.2, and b.3 below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection outage in an inspection period and the subsequent period begins at the conclusion of the included SG inspection outage.
- 1. After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period;
- 2. During the next 96 effective full power months, inspect 100% of the tubes.
This constitutes the second inspection period; and
- 3.
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.
Amendment Nos.
TS 6.4-13a
- c.
If crack indications are found in the portions of the SG tube not excluded above, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- 5. Provisions for monitoring operational primary to secondary LEAKAGE.
Amendment Nos.
TS 6.6-3
- b. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.1.D.4. In addition, the information itemized in Specification 3.1.D.4 shall be included in this report.
- 3. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after Tavg exceeds 200'F following completion of an inspection performed in accordance with the Specification 6.4.Q, Steam Generator (SG) Program. The report shall include:
- a. The scope of inspections performed on each SG,
- b. Degradation mechanisms found,
- c. Nondestructive examination techniques utilized for each degradation mechanism,
- d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e. Number of tubes plugged during the inspection outage for each degradation mechanism,
- f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator.
- g. The results of condition monitoring, including the results of tube pulls and in-situ testing,
- h. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, Amendment Nos.
4 4
TS 6.6-3a
- i. The calculated accident induced LEAKAGE rate from the portion of the tubes below 17.89 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced LEAKAGE rate from the most limiting accident is less than 1.80 times the maximum operational primary to secondary LEAKAGE rate, the report should describe how it was determined, and
- j.
The results of the monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
Amendment Nos.