NOC-AE-14003116, Application to Revise Technical Specifications to Adopt TSTF-510-A, Revision 2, Revision to Steam Generator Program Inspection Frequencies & Tube Sample Selection, Using the Consolidated Line Item Improvement Process

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Application to Revise Technical Specifications to Adopt TSTF-510-A, Revision 2, Revision to Steam Generator Program Inspection Frequencies & Tube Sample Selection, Using the Consolidated Line Item Improvement Process
ML15121A818
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 04/23/2015
From: Gerry Powell
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-14003116, STI 33849664
Download: ML15121A818 (21)
v • d • e


Text

Nuclear Operating Company South Texas Pro/ect ElectricGeneatin$ Station PO Box 289 Wadsworth. Texas 774S3 -

April 23, 2015 NOC-AE-14003116 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 Application to Revise Technical Specifications to Adopt TSTF-51 0-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Pursuant to 10CFR50.90, STP Nuclear Operating Company (STPNOC) is submitting a request for an amendment to the Technical Specifications (TS) for South Texas Project Units 1 and 2.

The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting based on Technical Specification Task Force (TSTF)-51 0-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involved no significiant hazards consideration. The bases for these determinations are included in the enclosure. In accordance with 10 CFR 50.91(b), STPNOC is notifying the State of Texas of this request for license amendment by providing a copy of this letter with the attachments. provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes. The changes to the TS Bases are provided for information only and will be implemented in accordance with TS 6.8.3.m, TS Bases Control Program, pursuant to approval of the license amendment.

The STPNOC Plant Operations Review Committee has reviewed and concurred with the proposed change.

STPNOC requests approval of the proposed amendment by December 1st, 2015 and requests 90 days from December 1st, 2015 for implementation of the amendment.

There are no regulatory commitments in this letter.

If there are questions regarding this submittal, please contact Hung C. Le at 361-972-7932, or me at 361-972-7566.

STI 33849664 A601 kJ/ L

NOC-AE-14003116 Page 2 of 3 I declare under penalty of perjury that the foregoing is true and correct.

Executed on kyýA 2-3 Z15 G. T. Powell Site Vice President ATTACHMENTS:

1. Evaluation of the Proposed Change
2. Technical Specification Page Markups
3. Technical Specification Bases Markups (for information only)

NOC-AE-14003116 Page 3 of 3 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV Steve Frantz U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner U. S. Nuclear Regulatory Commission Lisa. Regqner(anrc..qov Lisa M. Regner John Ragan Senior Licensing Project Manager Chris O'Hara U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8 B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Kevin Polio NRC Resident Inspector Cris Eugster U. S. Nuclear Regulatory Commission L. D. Blaylock P. 0. Box 289, Mail Code: MN1 16 CPS Energy Wadsworth, TX 77483 Attention: Document Control Desk U. S. Nuclear Regulatory Commission Peter Nemeth Washington, DC 20555-0001 Crain Caton & James, P.C.

C. Mele John Wester City of Austin Richard A. Ratliff Texas Department of State Health Services Robert Free Texas Department of State Health Services

NOC-AE-14003116 ATTACHMENT I Evaluation of the Proposed Change

Subject:

Application to Revise Technical Specifications to Adopt TSTF-51 0-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process

1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 REGULATORY EVALUATION

4.0 ENVIRONMENTAL EVALUATION

5.0 REFERENCES

NOC-AE-14003116 Attachment 1 Page 1 of 4 DESCRIPTION and ASSESSMENT

1.0 DESCRIPTION

1.1 Summary Description The proposed change revises South Texas Project (STP) Technical Specifications (TS) for STP Units 1 and 2 Limiting Condition for Operation (LCO) 3.4.5, "Steam Generator Tube Integrity,"

and Surveillance Requirement 4.4.5.2, and Administrative Controls Specification 6.8.3.o, "Steam Generator Program," and Specification 6.9.1.7, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications.

The proposed amendment is consistent with Technical Specification Task Force (TSTF) traveler TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sampling Selection." STP Units 1 and 2 replacement steam generators were constructed with Inconel 690TT tube material.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation STPNOC has reviewed TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection", and the model safety evaluation dated October 27, 2011. As described in the subsequent paragraphs, STPNOC has concluded that the justifications presented in TSTF-510-A, Revision 2, and the model safety evaluation prepared by the NRC staff are applicable to STP Units 1 and 2 and justify this amendment for the incorporation of the changes to the corresponding STP Units 1 and 2 TS.

2.2 Optional Chan-ges and Variations STPNOC is not proposing any significant variations or deviations from the TS changes described in TSTF-51 0, Revision 2, or the applicable parts of the NRC staff's model safety evaluation. Minor variations to the approved TSTF-510-A, Revision 2, are described below:

1. STP Units 1 and 2 are Custom Technical Specifications (TS) PWR plants and utilize different number than the Westinghouse Owners Group (WOG) Standard Technical Specifications (STS) on which TSTF-510 was based, that is Revision 3.1 of NUREG-1431, "Standard Technical Specifications for Westinghouse Plants." Specifically:

NOC-AE-14003116 Attachment 1 Page 2 of 4 These differences are administrative and do not affect the applicability of TSTF-510-A, Revision 2, to the STP Units 1 and 2 TS.

2. TSTF-51 0-A, Revision 2, Section 2.0, paragraph three, An editorial correction is made to Paragraph 5.5.9.b.1 (6.8.3.o.b.1 for STP Units 1 and 2 TS). The closing parenthesis is misplaced. It currently states "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, and all anticipated transients included in the design specification) and design basis accidents." This inappropriately includes anticipated transients in the description of normal operating conditions. The sentence is revised to, "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."
3. TSTF-51 0-A, Revision 2, Section 2.0, paragraph four, states "An editorial correction is made to add a missing closing bracket to the end of Paragraph 5.5.9.b.2." This change affects a bracketed (plant-specific) sentence in the WOG Standard TS. STP Units 1 and 2 did not incorporate the bracketed wording in corresponding TS 6.8.3.o and therefore this change does not apply.
4. The proposed change corrects an administrative inconsistency in TSTF-510-A, Revision 2, Paragraph d.2 of the Steam Generator Tube Inspection Program. In Section 2.0, "Proposed Change," TSTF-510-A states that references to "tube repair criteria" in Paragraph d is revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted as an administrative error (References 5.1 and 5.3):

"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).

STPNOC does not have an approved tube repair criteria. Therefore, the sentence is revised to state "tube plugging" criteria. In addition, all use of the phrase "tube repair' found in TSTF-51 0-A are omitted in the License Amendment Request (LAR). Thus "tube repair criteria" is replaced with "tube plugging criteria" throughout the LAR. For STP Units 1 and 2, the administrative error found in Paragrapg d.2 and related editorial corrections have been corrected in all affected paragraphs of the submitted markup for TS 6.8.3.o.

These changes are an administrative and should not result in this application being removed from the Consolidated Line Item Improvement Process.

In conclusion, these differences are administrative and do not affect the applicability of TSTF-510-A, Revision 2 to the STP TS.

NOC-AE-14003116 Attachment 1 Page 3 of 4

3.0 REGULATORY EVALUATION

3.1 No Significant Hazards Consideration Determination STPNOC requests adoption of an approved change to the standard technical specifications (STS) into the STP Units 1 and 2 plant specific technical specifications (TS), to revise LCO 3/4.4.5, "Steam Generator Tube Integrity," Surveillance Requirement 4.4.5.2, and Administrative Controls Specifications 6.8.3.o, "Steam Generator Program," and 6.9.1.7, "Steam Generator Tube Inspection Report," 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 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, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

NOC-AE-14003116 Attachment 1 Page 4 of 4 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.

Based on the above, STPNOC 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.

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.

5.0 REFERENCES

5.1 Technical Specifications Task Force TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," dated October 27, 2011 5.2 Federal Register Volume 76, Number 208, Page 66763 (76 FR 66763) dated October 27, 2011 5.3 Technical Specifications Task Force Correction to TSTF-51 0-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, "dated March 28, 2012

Attachment 2 Technical Specification Page Markups

NOC-AE-14003116 Attachment 2 Page 1 of 7 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Steam generator tube integrity shall be maintained.

AND All steam generator tubes satisfying the tube pluggin~ criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

NOTE: Separate entry is allowed for each steam generator tube

a. With one or more steam generator tubes satisfying the tube pluggingfopa.f criteria and not plugged in accordance with the Steam Generator Program, within 7 days verify tube integrity of the affected tube(s) is maintained until the next inspection, or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

AND Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or steam generator tube inspection.

b. With steam generator tube integrity not maintained, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVELLANCE REQUIREMENTS 4.4.5.1 Verify steam generator tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected steam generator tube that satisfies the tube pluggingfepaW criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a steam generator tube inspection.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-12 Unit 1 -Amendment No. 09, 107, 54., !*.

Unit 2 - Amendment No. 77, 94, 142,151

NOC-AE-14003116 Attachment 2 Page 2 of 7 6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.n (continued)

2) The ODCM shall also contain descriptions of the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and the Radiological Effluent Release Report required by Specifications 6.9.1.3 and 6.9.1.4.
3) Licensee-initiated changes to the ODCM:

a) Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1. Sufficient information to support the changes together with the appropriate analyses or evaluations justifying the changes and
2. A determination that the changes maintain the levels of radioactive effluent control required by 10 CFR 20.1 302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b) Shall become effective after approval of the plant manager.

c) Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (month and year) the change was implemented.

o. Steam Generator Program A Steam Generator Program shall be established and implemented to ensure that steam generator (SG) tube integrity is maintained. In addition, the Steam Generator Program shall include the following SOUTH TEXAS - UNITS 1 & 2 6-12 Unit I - Amendment No. 1-1,464 Unit 2 - Amendment No. 139, 454

NOC-AE-14003116 Attachment 2 Page 3 of 7 6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.o (continued)

a. 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.
b. Performance criteria for SG tube integrity. Steam generator tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.
1. Structural integrity performance criterion. All inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown), Gpd all anticipated transients included in the design specification,) and design basis accidents.

This includes retaining a safety factor of 3.0 (3AP) 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.

2. 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.

Accident induced leakage is not to exceed 1 gpm total for all four SGs in one unit.

SOUTH TEXAS - UNITS 1 & 2 6-12a Unit 1 - Amendment No. 4-64 Unit 2 - Amendment No. 4-4

NOC-AE-14003116 Attachment 2 Page 4 of 7 6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.o (continued)

3. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
c. Provisions for SG tube pluggingfepai 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
d. 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 pluggingrspai criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 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.
1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.eplaeemepA.
2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third 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 n each inspection period as defined in a, b, c and d 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 plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated.

NOC-AE-14003116 Attachment 2 Page 5 of 7 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.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

Inspect 100% of the tubes at sequential perieds of 1,44 108, 72, and, thereafter, 60 effective full power mFonths. The first sequential period shall be considered to begin after the ffirst ivie iof .Gs. the In addition, inspect 50% of the tubes by the refueling outage nearest the miedpoint of the period and the remaining 50% by the refueling outage nearest the end of the aeriod~ No SG shall onerate for More thann 72 toffiantiva full power months Or three refueli. Ia ouiaees !1WII*1eI without being inspected.

SOUTH TEXAS - UNITS 1 & 2 6-12b Unit 1 - Amendment No. 4-64 Unit 2 - Amendment No. 4-64

NOC-AE-14003116 Attachment 2 Page 6 of 7 6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.o (continued)

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially aff6dted 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 inspectionsis less). If definitive information, such as from examination of a pulled tube, diagnostic nondestructive 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.
e. Provisions for monitoring operational primary-to-secondary leakage.
p. Batterv Monitoring and Maintenance Program This Program provides for battery restoration and maintenance, which includes the following:
1) Actions to restore battery cells discovered with float voltage < 2.1 3 V;
2) Actions to equalize and test battery cells found with electrolyte level below the top of the plates;
3) Actions to verify that the remaining cells are > 2.07 V when a cell or cells are found to be < 2.13 V; AND
4) Actions to ensure that specific gravity readings are taken prior to each discharge test.
q. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Makeup and Cleanup Filtration System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
1. The definition of the CRE and the CRE boundary.
2. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

SOUTH TEXAS - UNITS 1 &2 6-12c Unit 1 - Amendment No. 164, 180, 185 Unit 2-Amendment No. 154, 167, 172

NOC-AE-14003116 Attachment 2 Page 7 of 7 6.0 ADMINISTRATIVE CONTROLS 6.9 Reporting Requirements 6.9.1.6 (continued)

c. The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided to the NRC upon issuance for each reload cycle.

6.9.1.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.3.o. The report shall include:

a. The scope of inspections performed on each SG,
b. ve4vDegradation 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 aetve degradation mechanism,
f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator-Total numberF and 196Mentage of tbe
g. The results of condition monitoring, including the results of tube pulls and in-situ testing, 6.9.2 Not Used SOUTH TEXAS - UNITS 1 & 2 6-17 Unit I -Amendment No.138, 14, 151, 164 Unit 2-Amendment No 127, 132, 139, 154

NOC-AE-14003116 Attachment 3 Technical Specifications Bases Page Markup (For Information Only)

NOC-AE-14003116 Attachment 3 Page 1 of 4 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY

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. SG 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.

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.

SG tubing is subject to a variety of degradation mechanisms. SG tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear.

These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.8.3.0, "Steam Generator Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.3.o, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. The SG performance criteria are described in Specification 6.8.3.o. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

Applicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary-to-secondary leakage rate equal to the operational leakage rate limits in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released via the main steam safety valves. The majority of the activity released to the atmosphere results from the tube rupture.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary-to-secondary leakage from all SGs of I gpm as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the limits in LCO 3.4.8, "Reactor Coolant System Specific Activity." For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose

NOC-AE-14003116 Attachment 3 Page 2 of 4 consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation (LCO)

The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the pluggingfep& criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugifopak criteria is removed from service by plugging. If a tube was determined to satisfy the pluggingfepu criteria but was not plugged, the tube may still have tube integrity. Refer to Action a.

below.

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 6.8.3.o 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." Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures that the primary-to-secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 1 gpm total from all SGs. The accident induced leakage rate includes any primary-to-secondary leakage existing prior to the accident in addition to primary-to-secondary leakage induced during the accident.

The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in LCO 3.4.6.2 and limits primary-to-secondary leakage through any one SG to 150 gpd. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress

NOC-AE-14003116 Attachment 3 Page 3 of 4 conditions of a LOCA or a main steam line break. Ifthis amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

Applicability Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4.

In MODES 5 and 6, primary-to-secondary differential pressure is low, resulting in lower stresses and reduced potential for leakage.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the required ACTIONS provide appropriate compensatory actions for each affected SG tube. Complying with the required ACTIONS may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated required ACTIONS.

a. The condition applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube pluggingfepaiF criteria but were not plugged in accordance with the Steam Generator Program as required by Surveillance Requirement 4.4.5.2. 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 pluggingrepa* 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 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, the plant must be shut down in accordance with the ACTION.

Seven days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, the ACTION statement allows plant 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 entering MODE 4 following the next refueling outage or SG inspection. This is acceptable since operation until the next inspection is supported by the operational assessment.

a. and b. Six hours to reach HOT STANDBY and an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to reach COLD SHUTDOWN are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Surveillance Requirements 4.4.5.1 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.

NOC-AE-14003116 Attachment 3 Page 4 of 4 A condition monitoring assessment of the SG tubes is performed during SG inspections. 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 pluggingiepai 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, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6).

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 6.8.3.o 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 6.8.3.o until subsequent inspections support extending the inspection interval.

4.4.5.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program pluggingrepai criteria is removed from service (by plugging). The tube pluggingrepei criteria delineated in Specification 6.8.3.0 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 pluggingfepai 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 6 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 entering MODE 4 following a SG inspection" ensures that the Surveillance has been completed and all tubes meeting the pluggingfepa4 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. 10 CFR 50 Appendix A, GDC 19
3. 10 CFR 100
4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB
5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976
6. EPRI Report, "Pressurized Water Reactor Steam Generator Examination Guidelines"
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