ML071200049

Tech Spec Pages for Amendments 233 and 228 Regarding Steam Generator Tube Surveillance Program

ML071200049
Person / Time
Site:	Turkey Point
Issue date:	04/27/2007
From:	NRC/NRR/ADRO/DORL/LPLII-2
To:
Mozafari B, NRR/ADRO/DORL, 415-2020
Shared Package
ML071080444	List: ML071080424 ML071200049
References
TAC MD1389, TAC MD1390
Download: ML071200049 (22)

Text

3 E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR 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 below:

A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).

B. Technical Specifications The Technical Sper-ifications contained in Appendix A, as revised through Amendment No. 233are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.

The licensee shall complete these activities no later than July 19, 2012.

The Final Safety Analysis Report supplement as revised on November 1, 2001,"

described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. DPR-31 Amendment No. 233

E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR 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 below:

A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).

B. Technical Specifications The Technical Sperifications contained in Appendix A, as revised through Amendment No. 228 are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on November 1,2001, describes certain future inspection activities to be completed before the period of extended operation.

The licensee shall complete these activities no later than July 19, 2012.

The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. DPR-41 Amendment No. 228

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation ......................................................................... 3/4 4-1 Ho t S ta n d b y ................................................................................................... 3/4 4-2 H o t S h u td ow n ................................................................................................. 3/4 4-3 C old S hutdow n - Loops Filled ......................................................................... 3/4 4-5 Cold Shutdown - Loops Not Filled .................................................................. 3/4 4-6 3/4.4.2 SAFETY VALVES S h u td ow n .............................................................................................. 3/4 4-7 O p e ra tin g ............................................................................................... 3/4 4-8 3/4.4.3 P R E S S UR IZ E R .............................................................................................. 3/4 4-9 3/4.4.4 R E LIE F VA LV E S ............................................................................................ 3/4 4-10 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY ............................................ 3/44-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage D etection S ystem s ........................................................................... 3/4 4-18 O pe ratio na l Lea kag e ...................................................................................... 3/4 4-19 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES ............ 3/4 4-22 3/4.4.7 C HE MIS T R Y .................................................................................................. 3/4 4-23 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ................................. 3/4 4-24 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE R E Q UIR E ME NT S ........................................................................................... 3/4 4-25 3/4.4.8 S P E C IF IC A C T IVIT Y ...................................................................................... 3/4 4-26 FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/gram DOSE EQUIVALENT 1-131 ............................................................................ 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS P R O G RA M ................................................................................................... 3/4 4-28 TURKEY POINT - UNITS 3 & 4 vii AMENDMENT NOS. 233 AND 228

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6 .6 DE LET ED .......................................................................................................................... 6 -12 6 .7 DE LET E D .......................................................................................................................... 6 -12 6.8 PROCEDURES AND PROGRAMS ................................... ............................................... 6-13 6.9 REPORTING REQ UIREM ENTS ........................................................................................ 6-18 6.9.1 ROUTINE REPO RTS .................................................................................................... 6-18 Startup Report ......................................................................................................... 6-18 Annual Reports ........................................................................................................ 6-19 Annual Radiological Environmental Operating Report ............................................ 6-20 Annual Radioactive Effluent Release Report .......................................................... 6-20 Peaking Factor Lim it Report .................................................................................... 6-21 Core Operating Limits Report .................................................................................. 6-21 Steam Generator Tube Inspection Report .............................................................. 6-22 6.9.2 SPECIAL REPO RTS .................................................................................................... 6-22a 6.10 DELETED ........................................................................................................................ 6-23 TURKEY POINT - UNITS 3 & 4 xvi AMENDMENT NOS. 233 AND 228

DEFINITIONS FREQUENCY NOTATION DOSE EQUIVALENT 1-131 1.12 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

E-AVERAGE DISINTEGRATION ENERGY 1.13 E shall be 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 (MeV/d) for the radionuclides in the sample isotopes, other than iodines, with half lives greater than 30 minutes, making up at least 95 percent of the total non-iodine activity in the coolant.

1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GAS DECAY TANK SYSTEM 1.15 A GAS DECAY TANK SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

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

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

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).

TURKEY POINT - UNITS 3 & 4 1-3 AMENDMENT NOS. 233 AND 228

DEFINITIONS OPERABLE - OPERABILITY 1.17 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, 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.

1.18 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 1.2.

PHYSICS TESTS 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (1) described in Chapter 13.5 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.20 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PURGE - PURGING 1.21 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

TURKEY POINT- UNITS 3 & 4 1-4 AMENDMENT NOS. 233 AND 228

TABLE 3.3-4 (Continued)

TABLE NOTAT1QION S During CORE ALTERATIONS or movement of irradiated fuel within the containment comply with Specification 3/4.9.13.

• • With irradiated fuel in the spent fuel pits.

1. Unit 4 Spent Fuel Pool Area is monitored by Plant Vent radioactivity instrumentation.

Note 1 Either the particulate or gaseous channel in the OPERABLE status will satisfy this LCO.

Note 2 Containment Gaseous Monitor Setpoint = (3.2 X 104) CPM, F )

Actual Purge Flow Where F = Design Purge Flow (35,000 CFM)

Setpoint may vary according to current plant conditions provided that the release rate does not exceed allowable limits provided in the Offsite Dose Calculation Manual.

ACTION STATEMENTS ACTION 26 - In MODES 1 thru 4: With both the Particulate and Gaseous Radioactivity Monitoring Systems inoperable, operation may continue for up to 7 days provided:

1) A Containment sump level monitoring system is OPERABLE,

2) Appropriate grab samples are obtained and analyzed at least once per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />,

3) A Reactor Coolant System water inventory balance is performed at least once per 8*** hours except when operating in shutdown cooling mode, and

4) Containment Purge, Exhaust and Instrument Air Bleed Valves are maintained closed.

Otherwise, be in at least HOT STANDBY within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br /> (ACTION 27 applies in MODES 5 and 6).

• • • Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after establishment of steady state operation.

TURKEY POINT - UNITS 3 & 4 3/4 3-37 AMENDMENT NOS. 233 AND 228

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 SG Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

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

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

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

b. With the requirements and associated allowable outage time of Action a above not met or SG tube integrity not maintained, be in HOT STANDBY within 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the next 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program.

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

Separate Action entry is allowed for each SG tube.

TURKEY POINT - UNITS 3 & 4 3/4 4-11 AMENDMENT NOS. 233 AND 228

THIS PAGE DELETED TURKEY POINT - UNITS 3 & 4 3/4 4-12 AMENDMENT NOS. 233 AND 228

THIS PAGE DELETED TURKEY POINT - UNITS 3 & 4 3/4 4-13 AMENDMENT NOS. 233 AND 228

THIS PAGE DELETED TURKEY POINT - UNITS 3 & 4 3/4 4-14 AMENDMENT NOS. 233 AND 228

THIS PAGE DELETED TURKEY POINT - UNITS 3 & 4 3/4 4-15 AMENDMENT NOS. 233 AND 228

THIS PAGE DELETED TURKEY POINT - UNITS 3 & 4 3/4 4-16 AMENDMENT NOS. 233 AND 228

THIS PAGE DELETED TURKEY POINT - UNITS 3 & 4 3/4 4-17 AMENDMENT NOS. 233 AND 228

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and

b. A Containment Sump Level Monitoring System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With both the Particulate and Gaseous Radioactivity Monitoring Systems inoperable, operation may continue for up to 7 days provided:

1) A Containment Sump Level Monitoring System is OPERABLE;

2) Appropriate grab samples are obtained and analyzed at least once per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />;

3) A Reactor Coolant System water inventory balance is performed at least once. per 8*

hours except when operating in shutdown cooling mode; and

4) Containment Purge, Exhaust and Instrument Air Bleed valves are maintained closed.

Otherwise, be in at least HOT STANDBY within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

b. With no Containment Sump Level Monitoring System operable, restore at least one Containment Sump Level Monitoring System to OPERABLE status within 7 days, or be in at least HOT STANDBY within 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection System shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Containment Sump Level Monitoring System-performance of CHANNEL CALIBRATION at least once per 18 months.

• Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after establishment of steady state operation.

TURKEY POINT - UNITS 3 & 4 3/4 4-18 AMENDMENT NOS. 233 AND 228

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATING 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4-1 up to a maximum of 5 GPM at a Reactor Coolant System pressure of 2235 +/- 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 .*

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding primary-to-secondary leakage, PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> or be in at least HOT STANDBY within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than allowed by 3.4.6.2.e above operation may continue provided:

1. Within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> verify that at least two valves in each high pressure line having a non-functional valve are in, and remain in that mode corresponding to the isolated condition, i.e., manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supplies deenergized. Follow applicable ACTION statement for the affected system, and

• Test pressure less than 2235 psig are allowed. Minimum differential test pressure shall not be less than 150 psid. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.

TURKEY POINT - UNITS 3 & 4 3/4 4-19 AMENDMENT NOS. 233 AND 228

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION (Continued)

2. The leakage* from the remaining isolating valves in each high pressure line having a valve not meeting the criteria of Table 3.4-1, as listed in Table 3.4-1, shall be determined and recorded daily. The positions of the other valves located in the high pressure line having the leaking valve shall be recorded daily unless they are manual valves located inside containment.

Otherwise be in at least HOT STANDBY within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

d. With any Reactor Coolant System Pressure Isolation Valve leakage greater than 5 gpm, reduce leakage to below 5 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or be in at least HOT STANDBY within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and in COLD SHUTDOWN within the following 30 hours1.25 days <br />0.179 weeks <br />0.0411 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity monitor at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.

b. Monitoring the containment sump level at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.

c.** Performance of a Reactor Coolant System water inventory balance at least once per 72*** hours; and

d. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />; and

e. Verifying primary-to-secondary leakage is < 150 gallons per day through any one SG at least once per 72*** hours.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage* to be within its limit:

a. At least once per 18 months.

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

• To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

• • Not applicable to primary-to-secondary leakage.

Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after establishment of steady state operation.

TURKEY POINT - UNITS 3 & 4 3/4 4-20 AMENDMENT NOS. 233 AND 228

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

The combined As-left leakage rates determined on a maximum pathway leakage rate basis for all penetrations shall be verified to be less than 0.60 La, prior to increasing primary coolant temperature above 200°F following an outage or shutdown that included Type B and Type C testing only.

The As-found leakage rates, determined on a minimum pathway leakage rate basis, for all newly tested penetrations when summed with the As-left minimum pathway leakage rate leakage rates for all other penetrations shall be less than 0.6 La, at all times when containment integrity is required.

3) Overall air lock leakage acceptance criteria is _<0.05 La, when pressurized to Pa.

The provisions of Specification 4.0.2 do not apply to the test frequencies contained within the Containment Leakage Rate Testing Program.

Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

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. 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 i.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).

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.

TURKEY POINT - UNITS 3 & 4 6-18 AMENDMENT NOS. 233 AND 228

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

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 SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and 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 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 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 total through all SGs and 500 gallons per day through any one 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.

The following alternate tube repair criteria may be applied as an alternative to the 40%

depth based criteria:

1. For Unit 3 during Refueling Outage 23 and the subsequent operating cycles until the next scheduled inspection, and for Unit 4 during Refueling Outage 23 and the subsequent operating cycles until the next scheduled inspection, flaws found in the portion of the tube below 17 inches from the top of the hot leg tubesheet do not require plugging.

For Unit 3 during Refueling Outage 23 and the subsequent operating cycles until the next scheduled inspection, and for Unit 4 during Refueling Outage 23 and the subsequent operating cycles until the next scheduled inspection, all tubes with flaws identified in the portion of the tube within the region from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged.

TURKEY POINT - UNITS 3 & 4 6-18a AMENDMENT NOS. 233 AND 228

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (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. For Unit 3 during Refueling Outage 23 and the subsequent operating cycles until the next scheduled inspection, and for Unit 4 during Refueling Outage 23 and the subsequent operating cycles until the next scheduled inspection, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded from inspection when the alternate tube repair criteria in Specification 6.8.4.j.c.1 is implemented. 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 tube 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 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 outages 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.

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-secondary leakage.

6.8.5 Administrative procedures shall be developed and implemented to limit the working hours of plant staff who perform safety-related functions, e.g. licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel. The procedures shall include guidelines on working hours that ensure that adequate shift coverage is maintained without routine heavy use of overtime for individuals.

Any deviation from the working hour guidelines shall be authorized by the applicable department manager or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant General Manager or his designee to assure that excessive hours have not been assigned. Routine deviation from the working hour guidelines shall not be authorized.

TURKEY POINT - UNITS 3 & 4 6-18b AMENDMENT NOS. 233 AND 228

ADMINISTRATIVE CONTROLS

3. WCAP-1 0054-P, Addendum 2, Revision 1 (proprietary), "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection in the Broken Loop and Improved Condensation Model," October 1995.*

4. WCAP-12945-P, "Westinghouse Code Qualification Document For Best Estimate LOCA Analysis," Volumes I-V, June 1996.**

5. USNRC Safety Evaluation Report, Letter from R. C. Jones (USNRC) to N. J. Liparulo (W),

"Acceptance for Referencing of the Topical Report WCAP-1 2945(P) 'Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Analysis,' " June 28, 1996.**

6. Letter dated June 13, 1996, from N. J. Liparulo (W) to Frank R. Orr (USNRC), "Re-Analysis Work Plans Using Final Best Estimate Methodology."

7. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," S. L. Davidson and T. L. Ryan, April 1995.

The analytical methods used to determine Rod Bank Insertion Limits and the All Rods Out position shall be those previously reviewed and approved by the NRC in:

1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

The ability to calculate the COLR nuclear design parameters are demonstrated in:

1. Florida Power & Light Company Topical Report NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point & St. Lucie Nuclear Plants."

Topical Report NF-TR-95-01 was approved by the NRC for use by Florida Power & Light Company in:

1. Safety Evaluation by the Office of Nuclear Reactor Regulations Related to Amendment No. 174 to Facility Operating License DPR-31 and Amendment No. 168 to Facility Operating License DPR-41, Florida Power & Light Company Turkey Point Units 3 and 4, Docket Nos. 50-250 and 50-251.

The AFD, FQ(Z), FAH, K(Z), and Rod Bank Insertion Limits shall be determined such that all applicable limits of the safety analyses are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector, unless otherwise approved by the Commission.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.8 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.j, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

• This reference is only to be used subsequent to NRC approval.

• • As evaluated in NRC Safety Evaluation dated December 20, 1997.

TURKEY POINT - UNITS 3 & 4 6-22 AMENDMENT NOS. 233 AND 228

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Cont'd)

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 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report as stated in the Specifications within Sections 3.0, 4.0, or 5.0.

TURKEY POINT - UNITS 3 & 4 6-22a AMENDMENT NOS. 233 AND 228