Text

Technical Specifications, Revising TS Steam Generator Tube Suveillance Program
	ML063110558
Person / Time
Site:	Fort Calhoun
Issue date:	11/07/2006
From:	Wang A; NRC/NRR/ADRO/DORL/LPLIV
To:	Ridenoure R; Omaha Public Power District
	Wang A, NRR/DORL/LP4, 415-1445
Shared Package
ML062780288	List: ML062780292; ML063110558;
References
	TAC MD2188
	Download: ML063110558 (30)
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(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 calibration or when 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 operation of the facility.

3.

This renewed 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 to the rules, regulations, and orders of the Commission now or hereafter In effect; and Is subject to the additional conditions specified or Incorporated below:

A.

Maximum Power Level Omaha Public Power District Is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not In excess of 1500 megawatts thermal (rated power).

B.

Technical Specifications The Technical Specifications contained In Appendix A, as revised through Amendment No., 246 are hereby !ncorporated In the license. Omaha Public Power District shall operate'the facility in accordance with the Technical Specifications.

C.

Security and Safeguards Contintiency Plans The Omaha Public Power District shall fully implement and maintain In effect all provisions of the Commis.sion-approvbd physical security, tram*ing and qualification, and safeguards contingency plans Including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 tind 10 CFR 50.54(p). The plans; which contain Safeguards Information'protected under.10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training And Quilification Plan, Safeguards Contingency Plan" submitted by letter dated October 18, 2004.

. ne.Oper.tiii-Lk-nse N.DPR-4 0

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 DELETED 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentation and Control Systems 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 Toxic Gas Monitors 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,,3,46,54,10,84,,6, 93,97,104,122,136,152, 160,176,183, 214,230,236, 246

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Not Used 5.6 Not Used 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Not Used 5.9.2 Not Used 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-Temperature Limits Report (PTLR) 5.10 Record Retention 5.11 Radiation Protection Program 5.12 DELETED 5.13 Secondary Water Chemistry 5.14 Systems Integrity 5.15 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.16.2 Radiological Environmental Monitoring Program 5.17 Offsite Dose Calculation Manual (ODCM) 5.18 Process Control Program (PCP) 5.19 Containment Leakage Rate Testing Program 5.20 Technical Specification (TS) Bases Control Program 5.21 Containment Tendon Testing Program 5.22 Diesel Fuel Oil Testing Program 5.23 Steam Generator (SG) Program 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED TOC - Page 3 Amendment No. 32,34,43,5.55,5-7, 236,-2g37 246

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE SECTION 1-1 R P S LS S S..................................................................................................................................

S ectio n 1.0 2-1 ESFS Initiation Instrumentation Setting. Limits.........................................................................

Section 2.14 2-2 Instrument Operating Requirements for RPS..........................................................................

Section 2.15 2-3 Instrument Operating Requirements for Engineered Safety Features.....................................

Section 2.15 2-4 Instrument Operating Conditions for Isolation Functions.........................................................

Section 2.15 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions........................ Section 2.15 2-9 RCS Pressure Isolation Valves.................................................................................................

Section 2.1 2-10 Post-Accident Monitoring, Instrumentation Operating Limits..................................................

Section 2.21 2-11 Toxic Gas Monitors Operating Limits.......................................................................................

Section 2.22 3-1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS.......................................

Section 3.1 3-2 Minimum Frequencies for Checks, Calibrations and Testing of................................................

Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks, Calibrations, and Testing....................................................

Section 3.1 of Miscellaneous Instrumentation and Controls 3-3a Minimum Frequency for Checks, Calibrations and Functional...................................................

Section 3.1 Testing of Alternate Shutdown Panels (AI-185 and AI-212) and Emergency Auxiliary Feedwater Panel (Al-1 79) Instrumentation and Control Circuits 3-4 Minimum Frequencies for Sampling Tests.................................................................................

Section 3.2 3-5 Minimum Frequencies for Equipment Tests...............................................................................

Section 3.2 3-6 Reactor Coolant Pump Surveillance..........................................................................................

Section 3.3 TOC - Page 4 Amendment No. 411,436,!69, 246

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE 5.2-1 M inim um Shift Crew Com position..............................................................................................

Section 5.0 TOC - Page 5 Amendment No. 116,125,142,115,152,476, 195, 246

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL ORDER)

Continued TABLE DESCRIPTION SECTION 2-10 Post-Accident Monitoring Instrumentation Operating Limits....................................................

Section 2.21 2-9 RCS Pressure Isolation Valves.................................................

................................................. Section 2.1 3-6 Reactor Coolant Pump Surveillance....................................................................................

...... Section 3.3 1-1 R P S LS S S..................................................................................................................................

S ection 1.0 2-11 Toxic Gas Monitoring Operating Limits....................................................................................

Section 2.22 TOC - Page 7 Amendment No. 116,15,152,17-6,195, 246

TECHNICAL SPECIFICATION DEFINITIONS

! Average Disintegration Energy is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

Offsite Dose Calculation Manual (ODCM)

The document(s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn/High (trip) Alarm setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain:

1)

The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2)

Descriptions of the information that should be included in the Annual Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports required by Specifications 5.9.4.a and 5.9.4.b.

Unrestricted Area Any area at or beyond the site boundary access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Core Operating Limits Report (COLR)

The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No. 1 specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal leakoff), that is captured and conductedto collection

.. systems or a sump or collecting tank,

2.

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

3.

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

Definitions - Page 8 Amendment No. 67,86,14!,152,164,224,226, 246 Correction letter of 06-17-2004

TECHNICAL SPECIFICATION DEFINITIONS

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal leakoff) that is not identified LEAKAGE, and

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

RCS Pressure-Temperature Limits Report (PTLR)

The PTLR is a fluence dependent document that provides Limiting Conditions for Operation (LCO) in the form of pressure-temperature (P-T) limits to ensure prevention of brittle fracture. In addition, this document establishes power operated relief valve setpoints which provide low temperature overpressure protection (LTOP) to assure the P-T limits are not exceeded during the most limiting LTOP event. The P-T limits and LTOP criteria in the PTLR are applicable through the effective full power years (EFPYs) specified in the PTLR. NRC approved methodologies are used as the bases for the information provided in the PTLR.

References (1) USAR, Section 7.2 (2) USAR, Section 7.3 Definitions - Page 9 Amendment No. 229, 246 Correction letter of 06-17-2004

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

(5)

DELETED (6)

Both steam generators shall be filled above the low steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300'F. Each steam generator shall be demonstrated operable by performance of the requirements specified in Section 3.17 prior to exceeding a reactor coolant temperature of 300'F.

(7)

Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia. A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8)

Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of the pressure and temperature limit Figure(s) shown in the PTLR.

(9)

Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature of 73 0F is required. Only 10 cycles are permitted.

(10)

Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of 73°F is required.

(11)

Low Temperature Overpressure Protection (LTOP)

(a)

The LTOP enable temperature and RCP operations shall be maintained in accordance with the PTLR.

(b)

The unit can not be placed on shutdown cooling until the RCS has cooled to an indicated RCS temperature of less than or equal to 300'F.

(c)

If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while Tr is below the LTOP enable temperature stated in the PTLR unless there is a minimum indicated pressurizer steam space of at least 50% by volume.

2.1 - Page 3 Amendment No. 39,56,66,71,119,136, 161,189, 207, 221, 246

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.4 Reactor Coolant System Leakage Limits Applicability Applies to the leakage rates of the reactor coolant system whenever the reactor coolant temperature (TcoId) is greater than 210 OF.

Obiective To specify limiting conditions of the reactor coolant system leakage rates.

Specifications To assure safe reactor operation, the following limiting conditions of the reactor coolant system leakage rates must be met:

(1)

RCS operational LEAKAGE shall be limited to:

a. No Pressure Boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE,

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

(2)

If RCS LEAKAGE limits of (1), above, are not met for reasons other than Pressure Boundary LEAKAGE or primary to secondary LEAKAGE, then reduce LEAKAGE to meet limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

(3)

If the Required Action and associated completion time of (2), above, is not met, OR Pressure Boundary LEAKAGE exists, or primary to secondary LEAKAGE, is not within limits, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(4)

To determine leakage to the containment, a containment atmosphere radiation monitor (gaseous or particulate) or dew point instrument, and a containment sump level instrument must be operable.

a. With no containment sump level instrument operable, verify that a containment atmosphere radiation monitor is operable, and restore the containment sump level instrument to operable status within 30 days.

b. With no containment atmosphere radiation monitor and no dewpoint instrument operable, restore either a radiation monitor or dewpoint instrument to operable status within 30 days.

c. With only the dewpoint instrument operable, or with no operable instruments, enter Specification 2.0.1 immediately.

2.1 - Page 13 Amendment No. 32,165,195, 226, 246

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.4 Reactor Coolant System Leakage Limits (Continued)

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS makeup system. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The 150 gallon per day operational limit on primary to secondary LEAKAGE through any one SG is based upon guidance in NEI 97-06, Steam Generator Program Guidelines. The Steam Generator Program operational LEAKAGE performance Criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY The potential for RCPB LEAKAGE is greatest when the RCS is pressurized, that is, when the reactor coolant temperature (Tcold) is greater than 210°F.

In MODES 4 and 5, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

REQUIRED ACTIONS (2).

Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to meet limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down.

This -action is necessary to prevent further deterioration of the RCPB.

REQUIRED ACTIONS (3)

If any pressure boundary LEAKAGE exists or primary to secondary LEAKAGE is not within limits, or if unidentified or identified LEAKAGE cannot be reduced to meet limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors

-that tend-to-de-grade-the pressure-boundary.-

The allowed Completion Times are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

2.1 - Page 16 Amendment No. 32,5*4

226, 246

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Intecqrity Applicability Applies whenever the reactor coolant temperature (Tjood) is greater than 210°F.

Obiective To ensure that SG tube integrity is maintained.

Specifications NOTE: Separate Condition entry is allowed for each SG Tube.

(1)

The following conditions shall be maintained:

(a)

SG tube integrity shall be maintained, and (b)

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

(2)

If one or more SG tube satisfy the tube repair criteria and are not plugged in accordance with the Steam Generator Program, then perform the following:

(a)

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

Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering MODE 4 following the next refueling outage or SG tube inspection.

(3)

If the Required Action and associated completion time of (2), above, is not met, or if SG tube integrity is not maintained, then be in MODE 3, Hot Shutdown, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Basis Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification-addresses-only-the-RCPB-integrity-funetion-of the-SG*T-he-SG-heat-removalfunction is-addressed by Technical Specification 2.1.1, "Operable Components."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

2.23-Page 1 Amendment No. 246

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Integrity (continued)

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 5.23, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.23, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.23. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

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

The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in Technical Specification 2.1.4. "Reactor Coolant System Leakage Limits," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes releases of activity occur from the faulted steam generator to the environment via the condenser air ejector and Main Steam Safety Valves (MSSVs) and Atmospheric Dump Valves (ADVs). The release via the condenser air ejector starts at the initiation of the event and continues to the reactor trip, while the release via the MSSVs/ADVs starts at the reactor trip and continues for the duration of the event.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the Technical Specification 2.1.3, "Reactor Coolant Radioactivity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

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

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.23, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

2.23-Page 2.

Amendment No. 246

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Inteqrity (continued)

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification,. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code, Section 111, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

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

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in Technical Specification 2.1.4. "Reactor Coolant System Leakage Limits," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

Steam generator tube integrity is challenged when the pressure differential across the tubes is large.

Large differential pressures across SG tubes can only be experienced in MODE 1, 2, or 3.

RCS conditions are far less challenging in MODES 4 and 5 than during MODES 1, 2, and 3. In MODES 4 and 5, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and appli*-6ti5Snof ssociate-d-Rquifed Ations.

Specification 2.23(2) applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by Technical Specification 3.17. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program.

2.23-Page 3 Amendment No. 246

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.23 Steam Generator (SG) Tube Integrity (continued)

The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Specification 2.23(3) applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action 2.23(2)b allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

If the Required Actions and associated Completion Times of Technical Specification 2.23(2) are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

References

1.

NEI 97-06, "Steam Generator Program Guidelines."

2.

10 CFR 50 Appendix A, GDC 19.

3.

10 CFR 100.

4.

ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes,"

August 1976.

6.

EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

2.23 - Page 4 Amendment No. 246

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Samplingq Tests (continued)

The Safety Injection (SI) pump room air treatment system consists of charcoal adsorbers which are installed in normally bypassed ducts. This system is designed to reduce the potential release of radioiodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers will assure system integrity and performance.

Pressure drops across the combined HEPA filters and charcoal adsorbers, of less than 9 inches of water for the control room filters (VA-64A & VA-64B) and of less than 6 inches of water for each of the other air treatment systems will indicate that the filters and adsorbers are not clogged by amounts of foreign matter that would interfere with performance to established levels. Operation of each system for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month will demonstrate operability and remove excessive moisture build-up in the adsorbers.

The hydrogen purge system provides the control of combustible gases (hydrogen) in containment for a post-LOCA environment. The surveillance tests provide assurance that the system is operable and capable of performing its design function. VA-80A or VA-80B is capable of controlling the expected hydrogen generation (67 SCFM) associated with 1)

Zirconium - water reactions, 2) radiolytic decomposition of sump water and 3) corrosion of metals within containment. The system should have a minimum of one blowerwith associated valves and piping (VA-80A or VA-80B) available at all times to meet the guidelines of Regulatory Guide 1.7 (1971).

If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.

Demonstration of the automatic and/or manual initiation capability will assure the system's availability.

Verifying Reactor Coolant System (RCS) leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained. Pressure boundary leakage would at first appear as unidentified leakage and can only be positively identified by inspection. Unidentified leakage is determined by performance of an RCS water inventory balance. Identified leakage is then determined by isolation and/or inspection. Since Primary to Secondary Leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance, note "***" for line item 8a on Tablee 3-5 states that the Reactor Coolant System Leakage surveillance is not applicable to Primary to Secondary Leakage. Primary to secondary leakage is measured by performance of effluent monitoring within the secondary steam and feedwater systems.

3.2 - Page 2 Amendment No. 45,67,!28,138,169, 246

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Sampling Tests (continued)

Table 3-5, Item 8b verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this surveillance requirement is not met, compliance with LCO 3.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and returns flows.

The Surveillance Frequency of daily is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

References

1)

USAR, Section 9.10

2)

ASTM D4057-95(2000), ASTM D975-98b, ASTM D4176-93, ASTM D129-00, ASTM D2622-87, ASTM D287-82, ASTM 6217-98, ASTM D2709-96

3)

ASTM D975-98b, Table 1

4)

Regulatory Guide 1.137

5)

EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

3.2 - Page 3b Amendment No. 22-9, 246

TECHNICAL SPECIFICATIONS

.TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS 1.

2.

3.

4.

Control Element Assemblies Control Element Assemblies Pressurizer Safety Valves Main Steam Safety Valves Test Drop times of all full-length CEA's Partial movement of all CEA's (Minimum of 6 in)

Verify each pressurizer safety valve is OPERABLE in accordance with the Inservice Testing Program.

Following testingq, lift settings shall be 2485 psig +/- 1 %/ and 2530 psig +/- 1%

respectively.

Frequency Prior to reactor criticality after each removal of the reactor vessel closure head USAR Section Reference 7.5.3 Q

R 7

7 Set Point 4

5.

DELETED

6.

DELETED

7.

DELETED

8.

Reactor Coolant E

System Leakage 8a.

Reactor CoolaInt E

System Leakage***

8b.

SG Tube Integ rity****

E 9a Diesel Fuel Su pply F

9b.

Diesel Lubricating Oil L

Inventory 9c.

Diesel Fuel Oil T

Properties 9d.

Required Diesel A

Generator Air Start Receiver Bank! Pressure valuate valuate valuate D

D*

D*

M 4

4 4

uel Inventory ube Oil Inventory est Properties 8.4 8.4 8.4 In accordance with the Diesel Fuel Oil Testing Program

.ir Pressure M

8.4 Amendment No. 15,24,128,160,166,169,171,219, 229, 246 3.2 - Page 6

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Whenever the system is at or above operating temperature and pressure.

Not applicable to primary to secondary LEAKAGE.

Verify primary to secondary LEAKAGE is _ 150 gallons per day through any one SG.

This surveillan e is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

3.2 - Page 7 Amendment No.

246

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Reference Test Frequency 9e.

1 Oa.

Check for and Remove Accum ulated Water from Each Fuel Oil Storage Tank Charcoal and HEPA Filters for Control Room Check for Water and Remove 1

In-Place Testinq**

Charcoal adsorbers and HEPA filter banks shall be leak tested and show >99.95%

Freon (R-11 or R-112) and cold DOP particulates removal, respectively.

Q On a refueling frequency or every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation or after each complete or partial replacement of the charcoal adsorber/HEPA filter banks, or after any major structural maintenance on the system housing or following significant painting, fire or chemical releases in a ventilation zone communicating with the system.

8.4 9.10

2.

Laboratory TestinQ**

Verify, within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows

methyliodide penetration less than 0.175% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (86°F) and a relative humidity of 70%.

- Tests shall be perforrned in accordance with applicable section(s) of ANSI N51C 3.2 - Pa On a refueling frequency or every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation or after any structural maintenance on the HEPA filter or charcoal adsorber housing or following significant painting, fire or chemical release in a ventilation zone communicating with the system.

1-1980.

ige 8 Amendment No. 15,24,128,169,19,,229, 246

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Reference Test Frequency 10a.

(continued)

3.

Overall System Operation

a.

Each circuit shall be operated.

Ten hours every month.

b.

The pressure drop across the R

combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 9 inches of water at system design flow rate.

c.

Fan shall be shown to operate R

within + 10% design flow.

4.

Automatic and manual initiation of the system shall be demonstrated.

R 10b.

Charcoal Adsorbers

1.

In-Place Testinq**

for Spent Fuel Charcoal adsorbers shall be Storage Pool Area leak tested and shall show

>99% Freon (R-1 1 or R-1 12) removal.

2.

Laboratory Testing Verify, within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows methyliodide penetration less than 10% when tested in accordance with ASTM D3803-1989 at a temperature of 300C (86°F) and a relative humidity of 95%.

- Tests shall be performed in accordance with applicable section(s) of ANSI N51 3.2 - PR On a refueling frequency or every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or after each.complete or partial replacement of the charcoal adsorber bank, or after any major structural maintenance on the system.

housing or following significant painting, fire or chemical release in a ventilation zone communicating with the system.

6.2 9.10 On a refueling frequency or every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation or after any structural maintenance on the HEPA filter or charcoal adsorber housing or following significant painting, fire or chemical release in a ventilation zone communicating with the system.

)-1980.

age 9 Amendment No. 15,24,52,128,151,1 69,198,22-9 246

TECHNICAL SPECIFICAT 1Ob.

(continued) 1Oc.

Charcoal Adsorlbers for S.I. Pump Room IONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test

3.

Overall System Operation

a.

Operation of each circuit shall be demonstrated.

b.

Volume flow rate through charcoal filter shall be shown to be between 4500 and 12,000 cfm.

4.

Manual initiation of the system shall be demon-strated.

1.

In-Place Testing**

Charcoal adsorbers shall be leak tested and shall show

>99% Freon (R-11 or R-112) removal.

2.

Laboratory Testinq Verify, within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows methyliodide penetration less than 10% when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (86'F) and a relative humidity of 95%.

U Frequency R

Ten hours every month.

R R

On a refueling frequency or every 9

720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or 6

after each complete or partial replacement of the charcoal adsorber bank, or after any major structural maintenance on the system housing or following significant painting, fire or chemical release in any ventilation zone communicating with the system.

On a refueling frequency or following 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation or after any structural maintenance on the HEPA filter or charcoal adsorber housing or following significant painting, fire or chemical release in a ventilation zone communicating with the system.

SAR Section eference

.10

.2

- Tests shall be perform~ed

3.

Overall System Operation

a.

Operation of each circuit Ten hours every month.

shall be demonstrated.

b.

Volume flow rate shall be R

shown to be between 3000 and 6000 cfm.

in accordance with applicable section(s) of ANSI N510-1980.

3.2 - Page 10 Amendment No. 15,24,52,128,169,198, 229 246

TECHNICAL SPECIFICATIONS Test TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Reference Frequency 1Oc.

(continued)

11.

Containment Ventilation System Fusible Linked Dampers

12.

Diesel Generatdr Calibrat Under-Voltage Fqelays

13.

Motor Operated Safety Injection Loop Valve Motor Starters (HCV-31 1, 314, 317, 320, 327, 329, 331, 333, 312, 315, 318, 321)

14.

Pressurizer Heaters i

15.

Spent Fuel Pool!

Racks

16.

Reactor Coolant Gas Vent Systei

4.

Automatic and/or manual initi-ation of the system shall be demonstrated.

1.

Demonstrate damper action.

2.

Test a spare fusible link.

te Verify the contactor pickup value at

<85% of 460 V.

R 1 year, 2 years, 5 years, and every 5 years thereafter.

9.10 R

R 8.4.3 Verify control circuits operation for post-accident heater use.

Test neutron poison samples for dimensional change, weight, neutron attenuation change and specific gravity change.

1.

Verify all manual isolation valves in each vent path are in the open position.

2.

Cycle each automatic valve in the vent path through at least one complete cycle of full travel from the control room. Verification of valve cycling may be determined by observation of position indicating lights.

3.

Verify flow through the reactor coolant vent system vent paths.

R 1, 2, 4, 7, and 10 years after installation, and every 5 years thereafter.

During each refueling outage just prior to plant start-up.

R R

3.2 - Page 11 Amendment No. 41,54,60,75,77,80,155,169,192,218,229 246

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test Frequency

17.

Hydrogen PurgE System

1.

Verify all manual valves are operable by completing at least one cycle.

2.

Cycle each automatic valve through at least one complete cycle of full travel from the control room. Verification of the valve cycling may be determined by the observation of position indicating lights.

3.

Initiate flow through the VA-80A and VA-80B blowers, HEPA filter, and charcoal adsorbers and verify that the system operates for at least (a) 30 minutes with suction from the auxiliary building (Room 59)

(b) 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with suction from the containment

4.

Verify the pressure drop across the VA-82 HEPAs and charcoal filter to be less than 6 inches of water. Verify a system flow rate of greater than 80 scfm and less than 230 scfm during system operation when tested in accordance with 3b. above.

1.

Verify required shutdown cooling loops are OPERABLE and one shutdown cooling loop is IN OPERATION.

2.

Verify correct breaker alignment and indicated power is available to the required shutdown cooling pump that is not IN OPERATION.

R R

a) M b) R R

18.

Shutdown Coolihig S (when shutdown cooling is required by TS 2.8).

W (when shutdown cooling is required by TS 2.8).

3.2 - Page 12 Amendment No. -38,1,8,, 8, 246

61 TECHNICAL SPECIFICATIONS Test TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Frequency

19.

Refueling Water Level

20.

Spent Fuel Pool Level

21.

Containment Penetrations

22.

Spent Fuel Ass4 Storage

23.

P-T Limit Curve

,ýmbly Verify refueling water level is >_ 23 ft. above the top of the reactor vessel flange.

Verify spent fuel pool water level is >_ 23 ft.

above the top of irradiated fuel assemblies seated in the storage racks.

Verify each required containment penetration is in the required status.

Verify by administrative means that initial enrichment and burnup of the fuel assembly is in accordance with Figure 2-10.

Verify RCS Pressure, RCS temperature, and RCS heatup and cooldown rates are within the limits specified by the P-T limit Figure(s) shown in the PTLR.

Verify by administrative means that initial enrichment and burnup of the fuel assembly is in accordance with Figure 2-11.

Prior to commencing, and daily during CORE ALTERATIONS and/or REFUELING OPERATIONS inside containment.

Prior to commencing, and weekly during REFUELING OPERATIONS in the spent fuel pool.

Prior to commencing, and weekly during CORE ALTERATIONS and/or REFUELING OPERATIONS in containment.

Prior to storing the fuel assembly in Region 2 (including peripheral cells).

This test is. only required during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

While these operations are occurring, this test shall be performed every 30 minutes.

Prior to placing the fuel assembly in a spent fuel cask in the spent fuel pool.

24.

Spent Fuel Cask Loading 3.2 - Page 13 Amendment No. 188, 221, 239 246

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes Applicability Applies to in-service surveillance of steam generator tubes.

Objective To ensure the integrity of the steam generator tubes.

Specifications Each steam generator shall be demonstrated OPERABLE by performance of the following:

(1)

Verify SG Tube Integrity in accordance with the Steam Generator Program at a frequency defined by the Steam Generator Program.

(2)

Verify that eachinspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to exceeding 21 0F reactor coolant temperature (Tcold).

(3)

Not Used.

(4)

Not Used.

(5)

Reportingq Requirements A report shall be submitted within 180 days after exceeding 210°F reactor coolant temperature (Tcold) following completion of an inspection performed in accordance with the Specification 5.23, Steam Generator (SG) Program. The report shall include:

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.

3.17 - Page 1 Amendment No. 404, 149, 246

TECHNICAL SPECIFICATIONS 3.17 - Page 2 through 3.17 - Page 9 not used Amendment No. 246

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued)

Basis During shutdown periods the SGs are inspected as required by this Surveillance Requirement (SR) and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

.The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e.,

which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.1 7(1). The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.23 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 5.23 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to exceeding 21 0°F reactor coolant temperature (Tcold) following a SG inspection ensures that the Surveillance has been completed and, all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

References

1.

NEI 97-06, "Steam Generator Program Guidelines."

2.

10 CFR 50 Appendix A, GDC 19.

3.-- --0-CFR -100.

4.

ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6.

EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

3.17 -Page 10 Amendment No. 04,195,228, 246

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.23 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 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, plugged, to confirm that the p5erformance 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, 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 in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.

3.

The operational LEAKAGE performance criterion is specified in LCO 2.1.4, "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.

5.0 - Page 19 Amendment No. 246

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.23 Steam Generator (SG) Program (continued)

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.

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.

5.0 - Page 20 Amendment No. 246

v t e Letter Sequence Other
TAC:MD2188, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Administration Questions Answers 30 May 2006 Results Approval, Approval, Approval, Approval Other: ML063110558
MONTHYEAR2006-11-07 [Table View]

ML063110558

Text

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