ML21252A758

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Supplement to License Amendment Request to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections
ML21252A758
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 09/09/2021
From: Schaeffer M
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID: L-2021-LLA-0151, NOC-AE-21003832, STI: 35207533, TSTF-577
Download: ML21252A758 (18)
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Text

September 9, 2021 NOC-AE-21003832 10 CFR 50.90 STI: 35207533 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 South Texas Project Units 1 & 2 Docket No. STN 50-498, 50-499 Supplement to License Amendment Request to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (EPID: L-2021-LLA-0151)

Reference:

Letter; J. Connolly to NRC Document Control Desk; Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections; August 10, 2021; (NOC-AE-21003812) (ML21222A227).

STP Nuclear Operating Company (STPNOC) requested approval of a license amendment to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections. STPNOC is providing this Supplement to correct editorial errors. These changes do not alter the technical content of the license amendment request.

STPNOC requests that pages 6-12c and 6-12d are removed from STP Units 1 and 2 Technical Specifications since these two pages would be blank. Pages 6-12e and 6-12f would be renumbered as 6-12c and 6-12d, respectively. The page renumbering is provided in the attached clean version of the Technical Specifications. It is not provided in the attached mark-up version.

There was an editorial error in mark-up copy of Technical Specification 6.9.1.6.b.11.

STPNOC requests that WCAP 124 72-P-A should be WCAP 12472-P-A. This error was not in the clean version of the Technical Specification. The correction is provided in the attached mark-up version of the Technical Specifications.

For Technical Specification 6.9.1.7.c.4, STPNOC requests that the bracket text or repaired be removed. STPNOC does not repair Steam Generator tubes. The removal is provided in the attached mark-up and clean versions of the Technical Specifications.

For new page 6-17b, STPNOC requests the page is renumbered to 6-17a. The change is provided in the attached mark-up and clean versions of the Technical Specifications.

All Technical Specification pages altered by the proposed license amendment request are included in this supplement for completeness.

There are no commitments in this letter.

If there are any questions or if additional information is needed, please contact Ali Albaaj at (361) 972-8949 or me at (361) 972-4778.

NOC-AE-21003832 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on 9/9 /o?^2/

Michael A. Schaeffer Site Vice President aa : Corrected Proposed Technical Specification Changes (Mark-Up) : Corrected Revised Technical Specification Pages ec:

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 E. Lamar Boulevard Arlington, TX 76011-4511 ec (electronic distribution):

Robert Free, Texas Department of State Health Services Dennis Galvin, Project Manager, U.S. Nuclear Regulatory Commission Gregory Kolcum, Senior Resident Inspector, U.S. Nuclear Regulatory Commission Chad Stott, Resident Inspector, U.S. Nuclear Regulatory Commission

NOC-AE-21003832 Corrected Proposed Technical Specification Changes (Mark-Up)

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 20.1302, 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 (SG) Program An SG Steam Generator Program shall be established and implemented to ensure that steam generator (SG) tube integrity is maintained. In addition, the SGSteam Generator Program shall include the following:

SOUTH TEXAS - UNITS 1 & 2 6-12 Unit 1 - Amendment No. 151, 164 209 Unit 2 - Amendment No. 139, 154 196

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. SGSteam 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), all anticipated transients included in the design specification, and design basis accidents.

This includes retaining a safety factor of 3.0 (3P) 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. 164 209 Unit 2 - Amendment No. 154 196

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 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.

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 plugging 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.

2.

After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 9672 effective full power months, which defines the inspection period. 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 in 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.

SOUTH TEXAS - UNITS 1 & 2 6-12b Unit 1 - Amendment No. 164, 209 Unit 2 - Amendment No. 154, 196

6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.o (continued)

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.

SOUTH TEXAS - UNITS 1 & 2 6-12c Unit 1 - Amendment No. 209 Unit 2 - Amendment No. 196

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 affected SG for the degradation mechanism that caused the crack indication shall be at the next 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 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.

Battery Monitoring and Maintenance Proqram This Program provides for battery restoration and maintenance, which includes the following:

1)

Actions to restore battery cells discovered with float voltage < 2.13 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 6-12d Unit 1 - Amendment No. 164, 180,185, 209 Unit 2 - Amendment No. 154, 167,172, 196

6.0 ADMINISTRATIVE CONTROLS 6.9 Reporting Requirements 6.9.1.6 (continued)

10.

WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997, (W Proprietary).

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient)

11.

WCAP 12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994 (W Proprietary), including Addenda 1-A (January 2000) and 4 (September 2012)

(Methodology for uncertainties in Specification 3.2.2-Heat Flux Hot Channel Factor and 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor)

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, Steam generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG;,

b.

The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c.

For each degradation mechanism found:

1.

The nondestructive examination techniques utilized;

2.

The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3.

A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4.

The number of tubes plugged during the inspection outage.

SOUTH TEXAS - UNITS 1 & 2 6-17 Unit 1 - Amendment No.138, 144, 151, 164, 204, 209, 213 Unit 2 - Amendment No 127, 132, 139, 154, 192, 196, 199

6.0 ADMINISTRATIVE CONTROLS 6.9 Reporting Requirements 6.9.1.7 (continued)

d.

An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

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, ef.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SGsteam generator.; and

f.

The results of any SG secondary side inspections.

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 6-17a Unit 1 - Amendment No.

Unit 2 - Amendment No

NOC-AE-21003832 Corrected Revised Technical Specification Pages

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 20.1302, 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 (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:
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. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

SOUTH TEXAS - UNITS 1 & 2 6-12 Unit 1 - Amendment No. 151, 164 209, Unit 2 - Amendment No. 139, 154 196,

6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.o (continued)

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), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 (3P) 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.
3. The operational leakage performance criterion is specified in LCO 3.4.6.2, Reactor Coolant System Operational Leakage.

C. 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.

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 plugging 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.

SOUTH TEXAS - UNITS 1 & 2 6-12a Unit 1 - Amendment No. 164, 209 Unit 2 - Amendment No. 154, 196

6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.o (continued)

2. After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. 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. Battery 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 < 2.13 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 6-12b Unit 1 - Amendment No. 164, 209, Unit 2 - Amendment No. 154, 196,

6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.q (continued )

3.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1) C.1.2 -- No peer reviews are required to be performed.
4.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by two trains of the Control Room Makeup and Cleanup Filtration System, operating at the flow rate required by the Surveillance Requirement 4.7.7.c.3, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

5.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph 3. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

6.

The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs 3 and 4, respectively.

SOUTH TEXAS - UNITS 1 & 2 6-12c Unit 1 - Amendment No.185 209 Unit 2 - Amendment No.172 196

6.0 ADMINISTRATIVE CONTROLS 6.8 Procedures, Programs, and Manuals 6.8.3.r Surveillance Frequency Control Program This program provides controls for surveillance frequencies. The program shall ensure that surveillance requirements specified in the technical specifications are performed at intervals sufficient to assure the associated limiting conditions for operations are met.

1)

The Surveillance Frequency Control Program shall contain a list of frequencies of those surveillance requirements for which the frequency is controlled by the program.

2)

Changes to the frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1.

STP takes the following exception to NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1:

a.

STP will use the Independent Decisionmaking Panel (IDP) described in the applications approved by the NRC for the Graded Quality Assurance Program and the Exemption from Certain Special Treatment Requirements, augmented by the Surveillance Test Coordinator and Subject Matter Expert(s), to perform the IDP function.

3)

The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

SOUTH TEXAS - UNITS 1 & 2 6-12d Unit 1 - Amendment No. 188 209 Unit 2 - Amendment No. 175 296

6.0 ADMINISTRATIVE CONTROLS 6.9 Reporting Requirements 6.9.1.6 (continued)

10.

WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997, (W Proprietary).

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient)

11.

WCAP 12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994 (W Proprietary), including Addenda 1-A (January 2000) and 4 (September 2012)

(Methodology for uncertainties in Specification 3.2.2-Heat Flux Hot Channel Factor and 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor)

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, "Steam Generator (SG) Program." The report shall include:

a.

The scope of inspections performed on each SG;

b.

The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c.

For each degradation mechanism found:

1.

The nondestructive examination techniques utilized;

2.

The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3.

A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4.

The number of tubes plugged during the inspection outage.

SOUTH TEXAS - UNITS 1 & 2 6-17 Unit 1 - Amendment No.138, 144, 151, 164, 204, 209, 213, Unit 2 - Amendment No 127, 132, 139, 154, 192, 196, 199,

6.0 ADMINISTRATIVE CONTROLS 6.9 Reporting Requirements 6.9.1.7 (continued)

d.

An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and

f.

The results of any SG secondary side inspections.

6.9.2 Not Used SOUTH TEXAS - UNITS 1 & 2 6-17a Unit 1 - Amendment No.

Unit 2 - Amendment No

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