ML063480167

Tech Spec Pages for Amendments 147 and 127, Tendon Surveillance Program

ML063480167
Person / Time
Site:	Vogtle
Issue date:	12/12/2006
From:	NRC/NRR/ADRO/DORL/LPLII-1
To:
Martin R, NRR/DORL, 415-1493
Shared Package
ML062970499	List: ML062970484 ML063480167
References
TAC MD1078, TAC MD1079
Download: ML063480167 (8)

Text

(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained inAppendix A,as revised through Amendment No. 147 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4) DELETED (5) DELETED (6) DELETED (7) DELETED (8) DELETED (9) DELETED (10) Additional Conditions The Additional Conditions contained in Appendix D,as revised through Amendment No. 102, are hereby incorporated into this license. Southern Nuclear shall operate the facility inaccordance with the Additional Conditions.

D. The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70.

These include (a) an exemption from the requirements of 10 CFR 70.24 for two criticality monitors around the fuel storage area, and (b) an exemption from the requirements of Paragraph Ill.D.2(b)(ii) of Appendix J of 10 CFR 50, the testing of containment air locks at times when containment integrity is not required. The special circumstances regarding exemption b are identified in Section 6.2.6 of SSER 5.

An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1 967, issued August 21, 1986, and relieved GPC from the requirement of having a criticality alarm system. GPC and Southern Nuclear are hereby exempted from the criticality alarm systern provision-oH0 CR70.24 so-far as-this section applies to the -

storage of fuel assemblies held under this license.

These exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. The exemptions in items b and c above are granted pursuant to 10CFR 50.12. With License No. NPF-68 Amendment No. 147

C. this license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and issubject to the additional conditions specified or incorporated below.

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained inAppendix A,as revised through Amendment No. 127 and the Environmental Protection Plan contained in Appendix B,both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 74. The SRs listed below shall be successfully demonstrated prior to the time and condition specified below for each:

a) DELETED b) DELETED c) SR 3.8.1.20 shall be successfully demonstrated at the first regularly scheduled performance after implementation of this license amendment.

(3) Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4) Additional Conditions The Additional Conditions contained inAppendix D,as revised through Amendment No. 80, are hereby incorporated into this license. Southern Nuclear shall operate the facility inaccordance with the Additional Conditions. -

D. The facility requires exemptions frdm certain requirements of 10 CFR Part 50 and 10 CFR Part 70.

These include (a) an exemption from the requirements of 10 CFR 70.24 for two criticality monitors around the fuel storage area, and (b) an exemption from the requirements of Paragraph lll.D.2(b)(ii) of Appendix J of 10 CFR 50, the testing of containment air locks at times when containment integrity isnot required. The special circumstances regarding exemption b are identified inSection 6.2.6 of SSER 8.

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TABLE OF CONTENTS LIST OF FIGURES 2.1.1-1 Reactor Core Safety Lim its ..................................................................... 2.0-2 3.4.16-1 Reactor Coolant Dose Equivalent 1-131 Reactor Coolant Specific Activity Limit Versus Percent of Rated Thermal Power with the Reactor Coolant Specific Activity > I pCi/gram Dose Equivalent 1-131 .................................. 3.4.16-4 3.7.18-1 Vogtle Unit I Burnup Credit Requirements for A ll C ell Storage .............................................................................. 3.7.18-3 3.7.18-2 Vogtle Unit 2 Burnup Credit Requirements for A ll C ell Storage .............................................................................. 3.7.18-4 4.3.1-1 Vogtle Units 1 and 2 Empty Cell Checkerboard Storage Configurations ................................................................... 4.0-6 4.3.1-2 Vogtle Unit 2 3x3 Checkerboard Storage Configuration ........................ 4.0-7 4.3.1-3 Vogtle Units 1 and 2 Interface Requirements (All Cell to Checkerboard Storage) ................................................. 4.0-8 4.3.1-4 Vogtle Unit 2 Interface Requirements (Checkerboard Storage Interface) .................. I............................. 4.0-9 4.3.1-5 Vogtle Unit 2 Interface Requirements (3x3 Checkerboard to All Cell Storage) .......................................... 4.0-10 4.3.1-6 Vogtle Unit 2 Interface Requirements (3x3 to Empty Cell Checkerboard Storage) .................................... 4.0-11 4.3.1-7 Vogtle Unit 1 IFBA Credit Requirements for All Cell Storage ......................................................................... ..... 4.0-12 4.3.1-8 Vogtle Unit 2 Bumup Credit Requirements for 3-out-of-4 Storage ................................. 4.0-13 4.3.1-9 Vogtle Unit 2 IFBA Credit Requirements for Center Assem bly for 3x3 Storage .............................................................. 4.0-14 4.3.1-10 Vogtle Unit 2 Bumup Credit Requirements for Peripheral Assemblies for 3x3 Storage .................................. 4.0-15 5.5.6-1 Deleted.

Vogtle Units l and 2 vi Amendment No. (Unit 1)

Amendment No. (Unit 2)

Amendment Nos. 147, 127

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Prestressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with ASME Boiler and Pressure Vessel Code Section XI, Subsection IWL and applicable addenda as required by 10 CFR 50.55a except where an exemption, relief, or alternative has been authorized by the NRC.

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.7 Reactor Coolant Pump Flywheel Inspection Proqram This program shall provide for the inspection of each reactor coolant pump flywheel at least once per 10 years by conducting either:

a. An in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius; or

b. A surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the disassembled flywheel.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program.

(continued)

Vogtle Units 1 and 2 5.5-5 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Amendment Nos. 147, 127

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.17 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Testing Program," dated September 1995, as modified by the following exceptions:

1. Leakage rate testing for containment purge valves with resilient seals is performed once per 18 months in accordance with LCO 3.6.3, SR 3.6.3.6 and SR 3.0.2.

2. Containment personnel air lock door seals will be tested prior to reestablishing containment integrity when the air lock has been used for containment entry. When containment integrity is required and the air lock has been used for containment entry, door seals will be tested at least once per 30 days during the period that containment entry(ies) is (are) being made.

3. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section Xl Code, Subsection IWL, except where relief or alternative has been authorized by the NRC. At the discretion of the licensee, the containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage.

4. A one time exception to NEI 94-01, Rev. 0, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix X:

Section 9.2.3: The next Type A test, after the March 2002 test for Unit 1 and the March 1995 test for Unit 2, shall be performed within 15 years.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, P,, is 37 psig.

The maximum allowable containment leakage rate, La, at Pa., is 0.2% of primary containment airweightper day.-.-- -7. ... .

(continued)

Vogtle Units I and 2 5.5-20 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Amendment Nos. 147, 127

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Deleted.

5.6.10 Steam Generator Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection. This report shall include:

1. Number and extent of tubes inspected,

2. Location and percent of wall thickness penetration for each indication of an imperfection, and

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 all be reported to the NRC within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

Vogtle Units I and 2 5.6-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Amendment Nos. 147, 127

Containment B 3.6.1 BASES (continued)

ACTIONS A.1 In the event containment is inoperable, containment must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when containment is inoperable is minimal.

B.1 and B.2 If containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and to MODE 5 within 36 hours1.5 days <br />0.214 weeks <br />0.0493 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. The containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage. The visual examinations of the steel liner plate inside containment are performed during maintenance or refueling outages since this is the only time the liner plate is fully accessible.

Failure to meet air lock and purge valve with resilient seal leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A, B, and C leakage causes that to exceed limits. Specific acceptance-criteria for--as-found and as-left leakage rates, as well as the methods of defining the leakage rates, are contained in the Containment Leakage Rate Testing Program.

At all other times between required leakage rate tests, the acceptance criteria are based on an overall Type A leakage limit of (continued)

Vogtle Units 1 and 2 B 3.6.1-4 Amendment Nos. 147, 127

Containment B 3.6.1 BASES SURVEILLANCE SR 3.6.1.1 (continued)

REQUIREMENTS

_ 1.OLa. At _*1.01La the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.

SR 3.6.1.2 For ungrouted, post-tensioned tendons, this SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program. Testing and Frequency are in accordance with ASME Boiler and Pressure Vessel Code Section Xl, Subsection IWL and applicable addenda as required by 10 CFR 50.55a except Where an exemption, relief, or alternative has been authorized by the NRC (Ref. 4).

REFERENCES 1. 10 CFR 50, Appendix J, Option B.

2. FSAR, Chapter 15.

3. FSAR, Section 6.2.

4. ASME Boiler and Pressure Vessel Code Section Xl, Subsection IWL and applicable addenda as required by 10 CFR 50.55a.

5. NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."

6. ANSI/ANS-56.8-1994, "American National Standard for Containment System Leakage Testing Requirement."

Vogtle Units 1 and 2 B 3.6.1-5 Amendment Nos. 147, 127