ML060550406

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Technical Specification, Revision of TS Associated with Steam Generator Tube Integrity Consistent with Revision 4 to Technical Specification Task Force, Standard TS Change Traveler, TSTF-449, Steam Generator Tube Integrity
ML060550406
Person / Time
Site: Sequoyah Tennessee Valley Authority icon.png
Issue date: 02/23/2006
From: Pickett D
Plant Licensing Branch III-2
To: Singer K
Tennessee Valley Authority
References
Download: ML060550406 (10)
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Text

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 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.ACTIONS*: a. With one or more SG tubes satisfying the tube repair 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 refueling outage or SG tube 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 />.ANbm b. Pluscthe affected tube(s) in accordance with the Steam Generator Program prior to startup followin~gsthe next refueling outage or SG tube inspection.

SURVE LLANCE REQUIREMENTS 4.4.5.0 Verify steam generator tube integrity in accordance with the Steam Generator Program.4.4.5.1 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to startup following a SG tube inspection.

MODES 1, 2, 3 and 4 ACTION: El. With any PRESSURE BOUNDARY LEAKAGE or with primary-to-secondary leakage not within limits, be in at least 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 following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.to. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE or primary-to-secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 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 following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.I I SURVEILLANCE REQUIREMENTS 6.9.1.16.1 Verify Reactor Coolant System lealkage is within limits by performance of a Reactor Coolant System 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 />.*The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.The above surveillance requirement is not applicable to primary-to-secondary leakage.4.4.6.2.2 Verify primary-to-secondary leakage is < 150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.** 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.

I I I SEQUOYAH -UNIT I 3/4 4-14 Amendment No. 12, 214, 222, 259, 306 ADMINISTRATIVE CONTROLS Technical Specification (TS) Bases Control Pronram This program provides a means for processing changes to the Bases of TSs.a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).k. Steam Generator (SG) ProQram 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 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 and/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.1. 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, cooldown, 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 SEQUOYAH -UNIT 1 6-11 Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306 6.0 ADMINISTRATIVE CONTROLS= = 2 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 accident induced leakage is not to exceed 1.0 gpm for the faulted SG. The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the maximum leakage rate established in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.3. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System, Operational Leakage.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 tubes at 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.

SEQUOYAH -UNIT 1 6-11la Amendment No. 306 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

6. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985, OA Proprietary)(Methodology for Specification 3/4.2.2 -Heat Flux Hot Channel Factor)7. WCAP-1 0266-P-A, Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, M Proprietary).(Methodology for Specification 3.2.2 -Heat Flux Hot Channel Factor).3. BAW-1 0227P-A, Evaluation of Advance Cladding and Structural Material (M5) in PWR Reactor Fuel," February 2000, (FCF Proprietary)(Methodology for Specification 3/4.2.2 -Heat Flux Hot Channel Factor)6.9.1.1 4.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Docum 3nt Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

specification 3.4.9.1, "RCS Pressure and Temperature (P/T) Limits" Specification 3.4.12, 'Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.1E .a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." 2. Westinghouse Topical Report WCAP-1 5293, "Sequoyah Unit I Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation." 3. Westinghouse Topical Report WCAP-1 5984, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Sequoyah Units 1 and 2." 6.9.1.15 b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.16 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.8.4.k, Steam Generator (EGG)Program. The report shall include: February 23, 2006 SEQUOYAH -UNIT 1 6-13a Amendment No. 52, 58, 72, 74, 117, 155, 223, 241, 258, 294, 297, 306 ADMINISTRATIVE CONTROLS a. The scope of inspections performed on each SG, 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, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and h. The effective plugging percentage for all plugging in each SG.SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.6.9.2.2 This specification has been deleted.6.10 R-CORD RETENTION (DELETED)SEQUOYAH -UNIT 1 6-14 Amendment No. 42, 52, 58, 72, 74, 117, 148, 155, 163, 174, 178, 223, 233, 241, 258, 294, 297, 306
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 leakage through a steam generator to the secondary system (primary to secondary leakage).MEMBER(S)

OF THE PUBLIC I 1.17 DELETED OFFSlrE DOSE CALCULATION MANUAL (ODCIA)1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used In the calculation of offsite doses resulting from radioactive gaseous and liquid effluent s, In the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and In the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluen: Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2)descriptions of the information that should be Included In the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.OPERABLE -OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILir(when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE -MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

SEQUOYAH -UNIT I 1-4 Amendment No. 12, 71, 148, 155, 169, 174, 178, 281., 306 PRESSURE BOUNDARY LEAKAGE 6.9 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage)through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)I 1.23 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the LTOP arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.15.PROCESS CONTROL PROGRAM (PCP)1.24 DELETED PURGE -PURGING 1.25 PURGE or PURGING is the controlled process of discharging air or gas from a confinement tc maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.RATED THERMAL POWER (RTP)1.27 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3455 MWt.REACTOR TRIP SYSTEM (RTS) RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its (RTS) trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by NRC.REPORTABLE EVENT 1.29 DELETED SEQUOYAH -UNIT 1 1-5 Amendment No. 12, 71, 141, 148, 155, 201, 233, 251, 275, 276, 294, 297, 306

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