ML071560531

Tech Spec Pages for Amendment 299 Steam Generator Tube Integrity

ML071560531
Person / Time
Site:	Millstone
Issue date:	05/31/2007
From:	NRC/NRR/ADRO/DORL/LPLI-2
To:
Hughey J, NRR/DORL, 301-415-3204
Shared Package
ML071380257	List: ML071380264 ML071560531 ML071560555
References
TAC MD2570
Download: ML071560531 (23)

Text

Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.'

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix, as revised through Amendment No. 299 , are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

Renewed License No. DPR-65 Amendment No. 299

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS Defined Term s ................................................................................................................ 1-1 Therm al Power ................................................................................................................ 1-1 Rated Therm al Pow er ....................................................................................................... 1-1 Operational M ode .............................................................. I............................................ 1-1 Action ............................................................................................................................... 1-1 Operable - Operability .................................................................................................... 1-1 Reportable Event .............................................................................................................. 1-1 Containm ent Integrity ...................................................................................................... 1-2 Channel Calibration ........................................................................................................ 1-2 Channel Check ...................... I .......................... 1-2 Channel Functional Test ................................................................................................... 1-2 Core Alteration ................................................................................................................ 1-3 Shutdown M argin ....................................................................................... ..................... 1-3 Leakage ............................................................................................................................ 1-3 Azim uthal Pow er Tilt .................................... .................................................................. 1-4 Dose Equivalent 1-131 ..................................................................................................... 1-4 E-Average D isintegration Energy .................................................................................... 1-4 Staggered Test Basis ......................................................................................................... 1-4 Frequency N otation ......................................................................................................... 1-4 Axial Shape Index ........................................................................................................... 1-5 Core Operating Lim its Report ........................................................................................ 1-5 MILLSTONE - UNIT 2 I Amendment No. 9, , 4-04, 444, 44-, 2 9 9

INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SA FETY VA LVES .................................................................................. 3/4 4-2 3/4.4.3 RELIEF VA LV E S .................................................................................... 3/4 4-3 3/4.4.4 PR E SSU R IZ E R ....................................................................................... 3/4 4-4 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY .......................................... 3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ...................................... 3/4 4-8 Leakage Detection Systems .................................................................... 3/4 4-8 Reactor Coolant System Operational Leakage ........................................ 3/4 4-9 3/4 .4.7 D E L E T E D ............................................................................................... 3/4 4-10 3/4.4.8 SPECIFIC ACTIVITY ............................................................................ 3/4 4-13 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ................................................. 3/4 4-17 R eactor Coolant System ........................................................................... 3/4 4-17 D E L E T E D ............................................................................................... 3/4 4-2 1 Overpressure Protection Systems ............................................................ 3/4 4-21 a 3/4 .4.10 D E L E T E D ............................................................................................... 3/4 4-22 3/4.4.11 D E L E T E D ............................................................................................... 3/4 4-23 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS ............................................................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - TavgŽ> 300°F ...................................................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 300°F .............................................. 3/45-7 3/4 5.4 REFUELING WATER STORAGE TANK ............................ ......... 3/4 5-8 3/4 5.5 TRISODIUM PHOSPHATE (TSP) ......................................................... 3/4 5-9 MILLSTONE - UNIT 2 VI Amendment No. -5, 72, 4-04, 4-53, -2 ,

264, 266,299

INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION ........... B 3/4.4-1 3/4.4.2 SA FETY VA LV ES ................................................................................... B 3/4 4-le 3/4.4.3 RELIEF VALVES ..................................... B 3/4 4-2 3/4.4.4 PRESSU RIZER ....................................................................................... B 3/4 4-2a 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY .......................................... B 3/4 4-2b 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ....................................... B 3/4 4-3 3/4.4.7 D E L ET E D .................................................................................................. B 3/4 4-4 3/4.4.8 SPECIFIC A CTIV ITY ............................................................................... B 3/4 4-4 3/4.4.9 PRESSURE/TEMPERATURE LIMITS .................................................... B 3/4 4-5 3/4.4.10 D E L ET E D ................................................................................................ B 3/4 4-7c 3/4.4.11 D E L E T E D .................................................................................................. B 3/4 4-8 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.*5.1 SAFETY INJECTION TANKS ................................................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSY STEM S ........................................................................ B 3/4 5-2 3/4.5.4 REFUELING WATER STORAGE TANK (RWST) ............................... B 3/4 5-2e 3/4.5.5 TRISODIUM PHOSPHATE (TSP) ........................................................... B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT .................................................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .............................. B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES .................... B 3/4 6-3b 3/4.6.4 COMBUSTIBLE GAS CONTROL ................................................ I.......... B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT ............................................................. B 3/4 6-5 MILLSTONE - UNIT 2 XII Amendment No. 66, 69, 7-2, 404, 4-53, 4-5, 247, 2-64, 2 6 6 ,2 9 1 ,'29-2,-293, 299

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ................. ................................................... 6-16 STARTUP REPORTS..................................................................... 6-16 ANNUAL REPORTS ..................................................................... 6-17 ANNUAL RADIOLOGICAL REPORT................................................. 6-18 CORE OPERATING LIMITS REPORT................................................. 6-18a STEAM GENERATOR TUBE INSPECTION REPORT ........................ I.......6-20 6.9.2 SPECIAL REPORTS...................................................................... 6-20

6. 10 DELETED 6.11 RADIATION PROTECTION PROGRAM............................................... 6-20a 6.12 HIGH RADIATION AREA................................................................ 6-20a 6.13 SYSTEMS INTEGRITY ................................................................... 6-23 6.14 IODINE MONITORING ................................................................... 6-23 6.15 RADIOLOGICAL EFFLUENT MONITORING AND OFESITE DOSE CALCULATION MANUAL (REMODCM')............................................ 6-24 6.16 RADIOACTIVE WASTE TREATMENT................................................. 6-24 6.17 SECONDARY WATER CHEMISTRY .................................................... 6-25 6.18 DELETED 6.19 CONTAINMENT LEAKAGE RATE TESTING PROGRAM ......................... 6-26 6.20 RADIOACTIVE EFFULENT CONTROLS PROGRAM ............................... 6-26 6.21 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ............... 6-28 MILLSTONE - UNIT 2 XVII Amendment No. -79,36,6-3,66,4-03, 4-104,44-4-,144-,4-5-3,4-6-3,1-69,239, 250, 62,264,240, 246, 2477,286, 294-, 29-,299

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.22 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM .................... 6-28 6.23 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM ....................... 6-28 6.24 DIESEL FUEL OIL TEST PROGRAM .......................................................................... 6-29 6.25 PRE-STRESSED CONCRETE CONTAINMENT TENDON SURV EILLAN CE PROGRA M .................................................................................... 6-29 6.26 STEAM GENERATOR PROGRAM ................................................................................ 6-30 MILLSTONE - UNIT 2 XVIII Amendment No. 2-7-8, 299

DEFINITIONS CORE ALTERATION 1.12 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

1.14.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals, and 1.14.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

1.14.3 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall, and 1.14.4 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all LEAKAGE which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

MILLSTONE - UNIT 2 1-3 Amendment No. , 2-63, 280, 299

REACTOR COOLANT SYSTEM STEAM GENERATOR TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Steam Generator (SG) tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

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

ACTION:

- - - - - - - - - --------- NOTE- - ----------------

Separate ACTION entry is allowed for each SG tube.

a. With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program:

1. Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection within 7 days, and

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

b. With required ACTION and associated completion time of ACTION a. not met or SG tube integrity not maintained:

1. 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

2. Be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

MILLSTONE - UNIT 2 3/4 4-5 Amendment No. 299

REACTOR COOLANT SYSTEM STEAM GENERATOR TUBE INTEGRITY SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program.

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

MILLSTONE - UNIT 2 3/4 4-6 Amendment No. 299

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 2 3/4 4-7 Amendment No. 2, -7, 5-2, 7-3, 83, 89, 4-0-+,

4-1-, 4--2-, 4-4&, 294--, 299

REMOVE THIS PAGE MILLSTONE - UNIT 2 3/4 4-7a Amendment No. 22, -7P,52, 89, 4-24, 3, 299

REMOVE THIS PAGE MILLSTONE - UNIT 2 3/4 4-7b Amendment No. -22,34, --, 89, 444, 2-9, 299

REMOVE THIS PAGE MILLSTONE - UNIT 2 3/4 4-7c Amendment No. 3-4,52- 299

REMOVE THIS PAGE MILLSTONE - UNIT 2 3/4 4-7d Amendment No.299

REMOVE THIS PAGE MILLSTONE - UNIT 2 3/4 4-7e Amendment No.299

REMOVE THIS PAGE MILLSTONE - UNIT 2 3/4 4-7f Amendment No. 2-2, 3-5,-2, 3, 89, 4-14, 29-1, 299

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System Operational LEAKAGE shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. I GPM UNIDENTIFIED LEAKAGE,

c. 75 GPD primary to secondary LEAKAGE through any one steam generator, and

d. 10 GPM IDENTIFIED LEAKAGE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any RCS operational LEAKAGE not within limits for reasons other than PRESSURE BOUNDARY LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. With ACTION and associated completion time of ACTION a. not met, or PRESSURE BOUNDARY LEAKAGE exists, or primary to secondary LEAKAGE not within limits, 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 be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1

---

NOTES -.-.-.-.----------------- -- --

1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

MILLSTONE - UNIT 2 3/4 4-9 Amendment No. 2-S, 37, 82, 8-5, 10,

+-24, 1-39, 24-, 228, 9& 299

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.2.2

---

NOTE -.----------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 75 gallons per day through any one SG at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

MILLSTONE - UNIT 2 3/4 4-10 Amendment No. 2-66, 299

ADMINISTRATIVE CONTROLS ANNUAL REPORTS 1 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted in accordance with 10 CFR 50.4 6.9.1.5a. DELETED 6.9.1.5b DELETED I 6.9.1.5c. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1)

Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3)

Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit. The report covering the previous calendar year shall be submitted prior to March 1 of each year.

I A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

MILLSTONE - UNIT 2 6-17 Amendment No. 3-6, 444, 44-S, 4-64, 2-70, 2-76, 2-86,299

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.9 A report shall be submitted within 180 days after initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.26, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SQ

b. Active 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 active degradation mechanism,

f. Total number and percentage of tubes plugged to date, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Deleted

b. Deleted

c. Deleted

d. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

e. Deleted

f. Deleted

g. RCS Overpressure Mitigation, Specification 3.4.9.3.

MILLSTONE - UNIT 2 6-20 Amendment No. 9, -36, 404, 444, 448, 4+Q, +-63,4-94, 239, 20, 266, 246, 248, 295, 299

ADMINISTRATIVE CONTROLS 6.9.2 (Continued)

h. Deleted

i. Tendon Surveillance Report, Specification 6.25

j. Deleted

k. Accident Monitoring Instrumentation, Specification 3.3.3.8.

1. Radiation Monitoring Instrumentation, Specification 3.3.3.1.

m. Deleted 6.10 Deleted.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and ( b ) of 10 CFR Part 20:

6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 tem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

MILLSTONE - UNIT 2 6-20a Amendment No. 29-S, 299

ADMINISTRATIVE CONTROLS 6.26 STEAM GENERATOR (SG) PROGRAM A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

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 a 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. Provisions for 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.

I. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including STARTUP, operation in the power range, HOT STANDBY, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

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 interms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 150 gpd per SG.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational LEAKAGE."

MILLSTONE - UNIT 2 6-30 Amendment No. 299

ADMINISTRATIVE CONTROLS 6.26 STEAM GENERATOR (SG) PROGRAM (Continued)

c. Provisions for SG tube repair 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 repair 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. An assessment of degradation 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 replacement.

2. Inspect 100% of the tubesat sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

MILLSTONE - UNIT 2 6-31 Amendment No. 299