ML062350198

Technical Specifications, Steam Generator Tube Integrity, MD0194 & MD0195

ML062350198
Person / Time
Site:	Point Beach
Issue date:	08/22/2006
From:	NRC/NRR/ADRO/DORL/LPLIII-1
To:
F. Lyon LPLE X2296
Shared Package
ML062050312	List: ML062050299 ML062350198
References
TAC MD0194, TAC MD0195
Download: ML062350198 (12)

Text

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.4 REACTOR COOLANT SYSTEM (continued) 3.4.5 RCS Loops - MODE 3..........................................................................

3.4.5-1 3.4.6 RCS Loops - MODE 4..........................................................................

3.4.6-1 3.4.7 RCS Loops - MODE 5, Loops Filled.....................................................

3.4.7-1 3.4.8 RCS Loops - MODE 5, Loops Not Filled..............................................

3.4.8-1 3.4.9 Pressurizer..........................................................................................

3.4.9-1 3.4.10 Pressurizer Safety Valves....................................................................

3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)...........................

3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP).............................

3.4.12-1 3.4.13 RCS Operational LEAKAGE................................................................

3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage......................................

3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation.............................................

3.4.15-1 3.4.16 RCS Specific Activity..........................................................................

3.4.16-1 3.4.17 Steam Generator (SG) Tube Integrity..................................................

3.4.17-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)...................................

3.5.1-1 3.5.1 Accumulators.......................................................................................

3.5.1-1 3.5.2 ECCS-Operating..................................................................................

3.5.2-1 3.5.3 ECCS-Shutdown..................................................................................

3.5.3-1 3.5.4 Refueling W ater Storage Tank (RW ST)...............................................

3.5.4-1 3.6 CONTAINMENT SYSTEMS.........................................................................

3.6.1-1 3.6.1 Containment........................................................................................

3.6.1-1 3.6.2 Containment Air Locks.........................................................................

3.6.2-1 3.6.3 Containment Isolation Valves...............................................................

3.6.3-1 3.6.4 Containment Pressure........................................................................

3.6.4-1 3.6.5 Containment Air Temperature..............................................................

3.6.5-1 3.6.6 Containment Spray and Cooling Systems...........................................

3.6.6-1 3.6.7 Spray Additive System.........................................................................

3.6.7-1 3.7 PLANT SYSTEMS........................................................................................ 3.7.1-1 3.7.1 Main Steam Safety Valves (MSSVs)....................................................

3.7.1-1 3.7.2 Main Steam Isolation Valves (MSIVs)

. and Non-Return Check Valves....................................................

3.7.2-1 3.7.3 Main Feedwater Isolation.....................................................................

3.7.3-1 3.7.4 Atmospheric Dump Valve (ADV) Flowpaths.........................................

3.7.4-1 3.7.5 Auxiliary Feedwater (AFW ) System.....................................................

3.7.5-1 3.7.6 Condensate Storage Tank (CST).........................................................

3.7.6-1 3.7.7 Component Cooling Water (CC) System.....................

3.7.7-1 3.7.8 Service W ater (SW ) System................................................................ 3.7.8-1 3.7.9 Control Room Emergency Filtration System (CREFS).........................

3.7.9-1 3.7.10 Fuel Storage Pool Water Level............................................................

3.7.10-1 3.7.11 Fuel Storage Pool Boron Concentration...............................................

3.7.11-1 3.7.12 Spent Fuel Pool Storage......................................................................

3.7.12-1 3.7.13 Secondary Specific Activity..............................

3.7.13-1 Point Beach 11 Unit 1 - Amendment No. 223 Unit 2 -Amendment No. 229

Definitions 1.1 1.1 Definitions LEAKAGE

-The maximum allowable primary containment leakage rate, La, shall be 0.4% of primary containment air weight per day at the peak design containment pressure (Pa).

LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to 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);

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c.

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

A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay.

The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

I MASTER RELAY TEST Point Beach 1.1-3 Unit 1 - Amendment No. 223 Unit 2 - Amendment No. 229

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

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

I APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

I Point Beach 3.4.13-1 Unit 1 - Amendment No. 223 Unit 2 - Amendment No. 229

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 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 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limits by performance of RCS water inventory balance.

SR 3.4.13.2 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 < 150 gallons per day through any one SG.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I

. Point Beach 3.4.13-2 Unit 1 -Amendment No.223 Unit 2 - Amendment No.229

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 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

1. Iftlt"Tlr I

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.

Program.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program.

next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

SURVEILLANCE REQUIREMENTS Point Beach 3.4.17-1 Unit I - Amendment No.223 Unit 2 - Amendment No.229

SG Tube Integrity

.3.4.17 SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program.

with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program.

a SG tube inspection Point Beach 3.4.17-2 Unit 1 - Amendment No.223 Unit 2 - Amendment No.229

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 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 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, 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 Point Beach 5.5-7 Unit 1 - Amendment No.

223 Unit 2 - Amendment No.

229

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 500 gallons per day per SG.

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS 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.

i. Unit 1 (alloy 600 Thermally Treated tubes): Inspect 100% of the tubes at sequential periods of 120, 90, 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 48 effective full power months or two refueling outages (whichever is less) without being inspected.

ii. Unit 2 (alloy 690 Thermally Treated tubes): 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 Point Beach 5.5-8 Unit 1 - Amendment No. 223 Unit 2 - Amendment No. 229

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Pro-gram (continued) 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.

Point Beach 5.5-9 Unit I - Amendment No.223 Unit 2 - Amendment No.229

Programs and Manuals 5.5 5.5 Programs and Manuals This page retained for page numbering Point Beach 5.5-10 Unit 1 - Amendment No. 223 Unit 2 - Amendment No. 229

Programs and Manuals 5.5 5.5 Programs and Manuals This page retained for page numbering Point Beach 5.5-11 Unit 1 - Amendment No. 223 Unit 2 - Amendment No. 229

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 5.6.8 Tendon Surveillance Report (continued)

Nuclear Regulatory Commission pursuant to the requirements of 10 CFR 50.4 within thirty days of that determination. Other conditions that indicate possible effects on the integrity of two or more tendons shall be reportable in the same manner. Such reports shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure and the corrective action taken.

Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4

.following completion of an inspection performed in accordance with the Specification 5.5.8, 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.

Point Beach 5.6-6 Unit 1 - Amendment No. 223 Unit 2 - Amendment No, 2 2 9