ML20086A031

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Proposed Tech Specs,Providing Revised Pages to Units 1 & 2 Tech Specs Based on Original Tech Spec Amend Pages Provided on 910226 & on NRC Approval of Reduced SG primary-to- Secondary Leakage Rate Submitted 911029
ML20086A031
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 11/13/1991
From:
ALABAMA POWER CO.
To:
Shared Package
ML19353B292 List:
References
NUDOCS 9111180198
Download: ML20086A031 (33)
v • d • e


Text

REACTOR COOLANT SYS1EM 3/4.4.6 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.6 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

EHQH:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 2000F.

SURVEILLANCE REQUIREMENTS 4.4.6.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.6.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.6.2 Steam Generator Tube # Samole Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.

The inservice inspectf on of steam generator tubes shall be performed at the frequencies sper.ified in Specification 4.1.6.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.6.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators.

Implementation of the steam generator tubc/ tube sup) ort plate elevation alternate plugging limit requires a 100% bobbin prose inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with outer diameter stress corrosion cracking (CD SCC) indications as determined by previous inspections and sampling. An inspection using a rotating pancake coil (RPC) probe is required for all tube support plate intersection indications exceeding a 1.5 volt bobbin coil signal amplitude. The RPC results are to be evaluated to establish if the principal indications can be characterized as OD SCC. Once an indication greater than 1.5 volts is characterized as OD SCC and left in service, RPC inspections of that indication will be performed at every second refueling outage.

When applying the exceptions of 4.4.6.2.a through 4.4.6.2.c, previous defects or imperfections in the area repaired by sleeving are~not considered an area requiring reinspection. The tubes selected for these inspections shall be selected on a random basis except:

1

  1. When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.6.4.a.9.

4 FARLEY - UNIT 1 3/4 4-9 AMENDMENT NO.

?111180199 91111.

IA ADOCK 0500 i

Attachment 2 Revised Technical Specification Pages Unit 1 PMLt 3/4 4-9 Replace 3/4 4-10 Replace 3/4 4-11 Replace 3/4 4-12 Replace 3/4 4-13 Replace 3/4 4-13a Insert 3/4 4-17 Replace B3/4 4-3 Replace B3/4 4-3a . Replace B3/4 4-4 Replace B3/4 4-5 Replace B3/4 4-5a Insert Unit 2 3/4 4-9 Peplace 3/4 4-10 Replace 3/4 4-11 Replace 3/4 4-12 Replace 3/4 4-13 Replace 3/4 4-13a Replace 3/4 4-13b Insert 3/4 4-17 Replace B3/4 4-3 Replace B3/4 4-3a Replace B3/4 4-4 Replace B3/4 4-5 Replace B3/4 4-Sa Insert

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

a. Where experience in similar slants with similar water chemistry indicates critical areas to se inspected, then at least 50% of the tubes inspected shall be from these critical areas.
b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1. All nonplugged tubes that previously had detectable wall penetrations greater than 20%.
2. Tubes in those areas where experience has indicated potential problems.
3. At least 3% of the total number of sleeved tubes in all three steam generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the tube and the sleeve.
4. A tube inspection (pursuant to Specification 4.4.6.4.a.8) shall be performed on each selected tube, if any selected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice irispection may be subjected to a partial tube inspection provided:
1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found. I
2. The inspections ir.clude thase portions of the tubes where  !

imperfections were previously found.

The results of each sample inspection shall be classified into one of the i following three categories Cateaorv Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

1 C-2 One or more tubes, but not more than 1% of the j total tubes inspected are defective, or between 5%  ;

and 10% of the total tubes inspected are degraded i tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected I tubes are defective.

FARLEY - UNIT 1 3/4 4-10 AMENDMENT NO. i

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Note: In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

4.4.6.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
b. If the results of the inservice inspection of a steam generator conducted in accordance.with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.6.3.a; the interval may then be extended to a maximum of once per 40 months.
c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.7.2.
2. A seismic occurriace greater than the Operating Basis Earthquake.
3. A loss-of-coolant accident requiring actuation of the engineered safeguards.
4. A main steam line or feedwater line break.
d. Tubes left in service in which the tube support plate elevation alternate plugging limit has been applied shall be inspected by bobbin coil probe during all future refueling outages.

FARLEY - UNIT 1 3/4 4-11 AMENDMENT NO.

} .- .__

REACTOR COOLANT SYSTEM I

SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.4 Acceotance Criteria ,

a. As used in this Specification:
1. Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal wall thickness, if detectable, may be considered as imperfections.
2. Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.  !

l

3. Deoraded Tube means a tube, including the sleeve if the tube l has been repaired, that contains imperfections greater than l or equal to 20% of the nominal wall thickness caused by )

degradation.

4.  % Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.
5. Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleevc containing a defect is defective.
6. Pluaoina or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e., sleeved) or removed fron service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of greater than or equal to 31% of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by piugging. For a tube that has been sleeved with a welded joint sleeve, through wall penetration greater than or equal to 37% of sleeve nominal wall thickness in the sleeve between the weld joints requires the tube to be removed from service by plugging. Also, this definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates. See 4.4.6.4.a.11 for the plugging limit for use within the thickness of the tube support plate.
7. Unserviceable describes the condition of a tube or sleeve if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.6.3.c, above.

FARLEY - UNIT 1 3/4 4-12 AMENDMENT NO.

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube that has been repaired by sleevir.g the tube inspection should include the sleeved portion of the tube. For a tube in which the tube support plate elevation alternate plugging limit has been applied, the bobbin inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of the last crack indication.
9. Tebe Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-ll178 Rev.1, or laser welded sleeving, as described by Westinghouse report WCAP-12672, which is used to maintain a tube in service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure.
10. P_tgigrvice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11. The Tube Support plate Alternate Pluaaina limit is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confineo within the thickness of the tube support plates. For application of the tube support plate alternate plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitnie. The plant specific guidelines used for eddy current inspections shall be amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the below listed voltage parameter. Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in the Farley steam generator inspections for consistent voltage normalization.
1. A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 4.0 volts, regardless of the depth of tube wall penetration, if, as a result, the projected end of cycle distribution of crack indications is verified to result in primary-to-secondary leakage less than 55 gpm in the faulted loop during a postulated steam line break event. The methodology for calculating expected leak rates from the projected crack distribution must be consistent with the latest NRC approved version of WCAP-12071.

FARLEY - UNIT 1 3/4 4-13 AMENDMENT N0.

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

2. A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 4.0 volts.
b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair of all tubes exceeding the plugging or repair limit) required by Table 4.4-2.

4.4.6.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission within 15 days of the completion of the plugging or repair effort.
b. The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1. Number and extent of tubes and sleeves inspected.
2. Location and percent of wall-thickness penetration for each indication of an imperfection.
3. Identification of tubes plugged or repaired.
c. Results of steam generator tubo inspections which fall into Category 4 C-3 shall be considered a rep 0RTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.
d. The resul+5 of inspections performed under 4.4.6.2 for all tubes left in service in which the tube support plate elevation alternate plugging limit has been applied shall be reported to the Commission following the inspection and prior to the resumption of plant operation. The report shall include:
1. Listing of applicable tubes.
2. Location (applicable intersections per tube) and extent of degradation (voltage).

FARLEY - UNIT 1 3/4 4-13a AMENDMENT NO.

1 BEACTOR COOLANT SYSTEM c OPERATIONAL LEAKAE LIMITING CONDIT10N FOR' OPERATION ,

3.4.7.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,
b. 1 GPM UNIDENTIFIED LEAKAGE,
c. 420 gallons per day total primary-to-secondary leakage through all steam generators and 140 gallons per day through any one steam generator,
d. 10 GPM IDENTIFIFD LEAKAGE from the Reactor Coolant System, and
e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 i 20 psig. .
f. The maximum allowable leakage of any Reactor Coolant System Pressure _ Isolation Yalve shall be as specified in Table 3.4-1 at a pressure of 2235 20 psig. 1 APPLICABILITY: MODES 1, 2, 3 and 4 1 ACTION:-
a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAYAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

-c. With any Reactor Coolant System Pressure Isolation Valve ,

' leakage greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from the. low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or.be in at

least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN ,

within the-following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS

........................s.................................................

4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by: >

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b. Monitoring the containment air cooler condensate level
system or containment atmosphere g
seous radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

FARLEY - UNIT 1 3/4 4-17 AMENDMENT NO. .

l REACTOR COOLANT SYSTEM  ;

BASES 3/4.4.6 STEAM CENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degrcdation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within-these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system. The allowable primary-to-secondary leak rate is 140 gallons per day per steam generator. Axial or circumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operational leakage of this magnitude can be readily detected by the existing Farley Unit I radiation monitors. Leakage in excess of this limit will require plant shutdown and an inspection, during which the leaking tubes will be located and plugged or repaired.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical leeve rl' 37% for the laser welded sleeve of sleeve nominal wall thickness in 4e sleeve, it must be plugged. The 31% and 37% limits are derived from R.G.

' 121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows:

FARLEY - UNIT 1 B3/4 4-3 AMENDMENT NO.

I REACTOR COOLANT SYSTEM  ;

l BASES  ;

a. Mechanical-Indications of degradation in the entire length of the sleeve 1.

must be evaluated against the sleeve plugging limit.

2. Indication of tube degradation of any type including a ,

complete guillotine break in the tube between the bottom of  !

the upper _ joint and the to) of the lower roll expansion does not require that the tube )e removed from service.

3. The tube plugging limit continues to apply to the aortion of the tube in the entire upper joint region and in tie lower '

roll expansion. As noted above the sleeve plugging limit applies-to these areas also.

4

4. The tube plugging limit continues to apply-to that portion of the tube above the top of the upper joint.

l

b. Laser Welded 3-
1. Indications of _ degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.

l 2. Indication of tube deoradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed

!- -from service.

3. At the weld joint, degradation must be evaluated in both the 4

sleeve and tube, 1 4 . In a joint with more than one weld, the weld closest to the i

end.of the sleeve represents the joint to be inspected and the limit of the sleeve inspection.

5. The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.

I- Steam generator tube inspections of operating plants have demonstrated

, the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.  ;

l  :

1- l I

)

i i l

' 1 FARLEY - UNIT 1 B3/4 4-3a AMENDMENT NO.  ;

l ki - _ . - - . _ , . . - _ , . . . _ _ _ _ . , _ . _ - , - . - . . - . - . _ -

REACTOR COOLANT SYSTEM BASES Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged ar repaired by the criterion of 4.4.6.4.a.ll.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to 10CFR50.73 prior to resumptiun of plant operation. Such cases will be considered by the Commission en a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, ' Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.7.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified partion c r this leakage can be reduced to a threshold value of less than 1 GPM. Th;s threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 31 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS Pressure Isolation valves is IDENTIFIED LEAKAGE and will be considered a portion of the allowed limit.

FARLEY - UNIT 1 B3/4 4-4 AMENDMENT NO.

, , , . . . , ,,-..n, e.,-. , - , .c-,. , , , - - , .

REACTOR COOLANT SYSTEM BASES The 140 GPD tube leakage limit per steam generator (420 GPD total) helps te maintain steam generator tube integrity in the event of a main steam line rupture or under LOCA conditions. By maintaining an operating leakage limit of 140 GPD per steam generator, the leak rate following a steam line rupture is also limited to 55 GPM in the faulted loop and 140 GPD per steam generator in the intact loops, which will limit offsite doses to within 10 percent of the 10 CFR 100 guidelines.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

3.4.4.8 CHEMISTRY The limitations on P.eactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

1 l

FARLEY - UNIT 1 B3/4 4-5 AMENDMENT N0.  :

l

REACTOR COOLANT SYSTEM BASES 3/4.4.9 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by t1e NRC of typical site locations. These values are conservative in that specific site parameters of the Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

As the leakage limit is restricted to 420 GPD total leakage (140 GPD per steam generator), offsite doses following a main steam line break are limited to 10 percent of the 10 CFR 100 guideline. This restriction is based on the results of a Farley site specific radiological evaluation that assumes a primary coolant iodine activity level corresponding to 1 percent fuel defects (approximately 4.0 microcurie / gram DOSE EQUIVALENT l-131) rather than a specific activity of 1.0 microcurie / gram DOSE EQUIVALENT I-131, and a post-accident primary-to-secondary leak rate of 55 GPM in the faulted loop.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

FAR!.EY - UNIT 1 B3/4 4-Sa AMENbMENT NO.

REACTOR COOLANT SYSTEM-3/4.4.6 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.6 fach steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ETl0H:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 2000F.

SURVEILLANCE REQUIREMENTS 4.4.6.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.6.1 Steam Generator Sample Selection and Inspn tion - Each steam generator shall be determined KrERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.6.2.1 Steam Generator Tube # Samole Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies. specified in Specification 4.4.6.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.6.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators. Selection of tubes to be inspected-is not affected by the F* designation, implementation of the steam generator tube / tube support plate elevation alternate plugging limit requires _a 100% bobbin probe inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with-outer diameter stress corrosion cracking (OD SCC) indications as determined by-previous inspections and sampling. An inspection using the rotating pancake coil (RPC) probe is required for all tube support plate intersection indications exceeding a 1.5 volt bobbin coil signal amplitude.

The RPC-results are to be evaluated to establish if the principal indications can be characterized as OD SCC. Once an indication greater than 1.5 volts is characterized as OD SCC and left in service, RPC inspections of that indication will be performed at every_second refueling outage.

When applying the exceptions of 4.4.6.2.1.a through 4.4.6.2.1.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection. The tubes selected for these inspections shall be i

selected on a random basis except:

  1. When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.6.4.a.9.

FARLEY - UNIT 2 3/4 4-9 AMENDMENT NO.

REACTOR COOLANT SYSTEM

- SVRVEILLANCE REQUIREMENTS (Continued)

a. Where experience _in similar plants with similar water chemistry '

indicates critical areas to be inspected, then at least 50% of the tubes. inspected shall be from these critical areas.

b. The-first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1. - All nonplugged tubes-that previously had detectable wall penetrations greater than 20%.
2. Tubes in those areas where experience has indicated potential.

problems.

3. At least 3% of the' total number of sleeved tubes in all three stea:a generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is.less. These inspections will_ include both the tube and the sleeve.
4. A tube inspection (pursuant to Specification 4.4.6.4.a.8) shall be performed on each selected tube. If any. selected tube does not permit the passage of the eddy. current probe for a tube or sleeve inspection, this shall be recorded and an adjacent-tube shall be selected and subjected to a tube-

~

inspection.

I c. The tubes-selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to

[ a partial tube inspection provided:

!' 1. The tubes. selected for these samples include the tubes from those areas of the tube sheet array.where tubes with l imperfections were previously found.

2. The inspections include those portions of the tubes where ,

imperfections were previously found.

~

o i

i l

i

$=

FARLEY - UNIT 2 3/4 4-10 AMENDMENT NO.

4

,.,-.-,,,..._..--,_,-s.,_m , ,__~ _ . . . _ --,,-. - ~ -. __. ,,,.. , ,.~,- - , , _ _ . - . -_

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

........................ ....................................u............

The results of each sample inspection shall be classified into one of the following three categories:

Cateoory Insoection Results C-1 Less thar. 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes in m eted are defective, or between 5% and 10% of tie fctai tubes inspected are y degradd tubes.

C-3 More than 10% of the total tubes inspected are degraderi tuber or more than 1% of the inspected tubes are defective.

Note: In all inspections, pr3viously degraded tubes or slaaves must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage

  • calculations.

4.4.6.2.2 Steam Generator F* Tube Inspection - In addition to the minimum sample size as determined by Specification 4.4.6.2.1, all F* tubes will be inspected within the tubesheet region. The results of this inspection will not be a cause for addi+ional inspections per Table 4.4-2.

i

, 1.4.6.3 Insoection Frecuencies - The above required inservice inspections ',

of steam generator tubes c!,all be performed at the following frequencies:

i a. The first inservice inspection shall be performed after 6

~

ET 1ctive Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspectian. If two consecutive inspecticra f e m ing service under AVT conditions, not including

, the preserv .t .cction, result in- all inspection results falling int 3 tw .-l category or if two cor.,ecutive inspections demonstrate cnat previously observed degradation has not continued

, and no additional degradation has occurreo, the inspe: tion interval may be extended to a maximum of once per 40 months.

FARLEY - UNIT 2 3/4 4-11 AMENDMENT NO.

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

....................y.........................................................

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months The increase in inspection frequency shall apply until the subsequent inspections satisfy the critoria of Specification 4.4.6.3.a; the interval may then be extended to a maximum of once per 40 months.
c. Additional, unscheduled inservice intpections shall be performed on each stea:n generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1. Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube :;heet welds) in excess of the limits of Specification 3.4.7.2.
2. A seismic occurrence greater than the Operating Basis Eart hquake.
3. A loss-of-coolr* accident requiring actuation of the engineered safeguards. ,
4. A main steam line or feedwater line break.
d. Tubes left in service in which the tube support plate elevation alternate plugging limit has been applied shall be inspected by bobbin coil probe during all future refueling outages.

( 4.6.4 Acceptance Criteria

a. As used in this Specification: _
l. hL .-fection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal wall thickness, if detectable, may be considered as imperfections.

S.

Dearadation mear.s a service-induced cracking,

wastage, wear or general corrosion occurring on either inside or outside of a s tube or sleeve.

FARLEY - UNIT 2 3/4 4-12 AMENDMENT NO.

1

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

3. D.qgaded Tube means a tube, including the sleeve if the tube has been repaired, that contains imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.  % Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.
5. Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleeve containing a defect is

, defective.

4 6. Pluaaino or Reoair Limit means the irperfection depth at or beyond which the tube shall be repaired (i.0., sleeved) or removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. This definition does not apply to the area of the tubesheet region below the F* distance in F*

tubes. For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of greater than or equal to 31%

of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging. For a tube that has been sleeved with a welded joint sleeve, through wall penetration greater than or equal to 37% of sleese nominal wall thickness in the sleeve between the weld joints requires the tube to be removed '

from service by plugging. This definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates. See 4.4.6.4.a.14 for the plugging limit for use within the thickness of the tube support plate.

7. Unserviceable describes the condition of a tube or sleeve if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.6.3.c, above.
8. Tube Inspection means an inspection of the steam generator tube from the peint of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of the tube. For a tube in which the tube support plate elevation alternate plugging limit has been applied, the
bobbin inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of the last crack indication.

l 3

FARLEY - UNIT 2 3/4 4-13 AMENDMENT NO.

BEACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

9. Tube Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-11178 Rev.1, or laser welded sleeving, as described by Westinghouse report WCAP-12672, which is used to maintain a tube in service or return a tube to service. This includes the removal of plugs that were installed as a corrective or praventive measure.
10. Preservice Insoection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11. F* Distance is the distance of the expanded portion of a tube which provides a sufficient length of undegraded tube expansion to resist pullout of the tube from the tubesheet. The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.
12. F* Tube is a tube:

a) with degradation equal to or greater than 40% below the F* distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains in service.

13. Tube Excansion is that portion of a tube which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the hole in the tu'oesheet.
14. The Tube Support Plate Alternate Pluaaina limit is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates. For application of the tube support plate alternate plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude. The plant specific guidelines used for eddy current inspections shall be amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the below listed voltage parameter. Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in the Farley steam generator inspections for consistent voltage normalization. l l

FARLEY - UNIT 2 3/4 4-13a AMENDMENT N0. l 1

l

REACTOR COOLANT SYSTEM

. SURVEILLANCE REQUIREMENTS _(Continued)-

.l. A tube caa remain in service. if the signal amplitude of a  :

crack indication .is.less than or equal to 4.0 volts, regardless of the depth of tube wall penetration, if, as a result, the projected end of cycle distribution of crack

-indications _ is verified to result in primary-to-secondary leakage less than' 55 gpm in the f.aulted loop during a postulated steam line break event'. The methodology for calculating expected leak rates from the projected crack distribution must be consistent with:the-latest NRC approved version of WCAP-12871. ,

2. A tube should be plugged or repaired if the signal amplitude of the crack indication _is greater than 4.0 volts,
b. The steam generator shall be determined 0PERABLE after completing the corresponding actions (plug or_ repair of all tubes exceeding the plugging.or repair limit) required by Table 4.4-2.

4.4.6.5 Reports

a. Following.each inservice inspection of steam generator tubes,'the number of tubes plugged, repair'ed or designated F* in each steam generator shall be reported to the Commission within 15 days of the completion of the inspection, plugging or repair effort.
b. The complete results of the steam generator tube and sleeve inservice inspection shall-be submitted to the Commission in _a Special Report pursuant to Specification 6.9.2 within 12 months following the ,

completion of the inspection. This Special Report shall include:

1. Number and extent of tubes and sleeves-inspected.
2. Location and percent of wall-thickness penetration for each indication of an imperfection.
3. Identification of tubes plugged or repaired.

Results of steam generator tube inspections which fall into Category c.

C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior _to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

d. The results of inspections performed under 4.4.6.2_for all tubes left in service in which the tube support plate elevation alternate plugging limit has been applied shall be reported to the Commission following the inspection and prior to the resumption of plant operation. The report shall include:
1. Listing of applicable tubes.
2. Location (applicable intersections per tube) and extent of degradation (voltage).

FARLEY - UNIT 2 3/4 4-13b AMENDMENT NO.

I

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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited .to:

a. No PRESSURE BOUNDARY LEAKAGE,
b. 1 GPM UNIDENTIFIED LEAKAGE,

.c. 450 gallons per day total primary-to-secondary leakage through all steam ger.erators and 150 gallons per day through any one steam generator,-

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 i-20 psig,
f. The maximum allowable leakage of any Reactor Coolant System cressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 2235 i 20 psig.

-APPLICABILITY: MODES-1, 2, 3 and=4 ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within.6' hours and in COLD SHUTDOWN within the following 30
hours.
b. With any Reactor Coolant-System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE, reduce-the leakage rate-to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the fo11owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c. With any Reactor Coolant System Pressure Isolation Valv'e-leakage greater-than the limit specified in. Table 3.4-1, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed-manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

~ SURVEILLANCE REQUIREMENTS i

4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

1

, a. Monitoring the containment atmosphere particulate radioactivity '

moniter at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  ;

1

b. Monitoring the containment air cooler condensate level system or l containment atmosphere gaseous radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
FARLEY - UNIT 2 3/4 4-17 AMENDMENT NO. j l

REACTOR COOLANT SYSTEM BASES 3/4.4.6 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system. The allowable primary-to-secondary leak rate is 150 gallons per day per steam generator.

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operational leakage of this magnitude can be readily detected by the existing Farley Unit 2 radiation monitors. Leakage in excess of this limit will require plant shutdown and an inspection, during which the leaking tubes will be located and plugged or repaired.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 37% for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R.G. 1.121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows:

a. Mechanical
1. Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.

FARLEY - UNIT 2 B3/4 4-3 AMENDMENT N0.

REACTOR COOLANT SYSTEM BASES

2. Indication of tube degradation of any type including a complete guillotine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.
3. The tube plugging limit continues to apply to the portion of the tube in the entire upper joint region and in the lower roll expansion. As noted above the sleeve plugging limit applies to these areas also.
4. The tube plugging limit continues to apply to that portion of the

' tube above the top of the upper joint.

b. Laser Welded
1. Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.
2. Indication of tube degradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed from service.
3. At the weld joint, degradation must be evaluated in both the sleeve and tube.
4. In a joint with more than one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the limit of the sleeve inspection.
5. The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.

F* tubes do not have to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded. The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation. Included in this distance is an allowance of 0.25 inch for uddy current elevation measurement uncertainty.

l Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criterion of 4.4.6.4.a.14.

l Steam generator tube inspections of operating plants have demonstrated

l. the capability to reliably detect wastage type degradation that has l

penetrated 20% of the original tube wall thickness.

l FARLEY - UNIT 2 B3/4 4-3a AMENDMENT NO.

l

1 REACTOR COOLANT SYSTEM l BASES

.............e............................................................

Whenever the results of any steam generator tubing inservice inspection I fall into Category C-3, these results will be reported to the Commission pursuant to 10CFR50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if j necessary.

3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.7.2 OPERATIO"AL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 31 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less thin assumed in the accident analyses.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered a portion of the allowed limit.

4

(

l FARLEY - UNIT E B3/4 4-4 AMENDMENT N0.

w

REACTOR COOLANT SYSTEM BASES

............................................m..............................

The 150 GPD tube leakage limit per steam generator (450 GPD total) helps to maintain steam oenerator tube-integrity in tie event of a main steam line rupture or under LOCA conditions. By maintainit., an operating leakage limit of 150 GPD per steam generator, the leak rate following a steam line rupture is also limited to 55 GPM in the faulted loop and ISO GPD per steam gen 6rator in the intact loops, which will limit offsite doses to within 10 percent of the 10 CFR 100 guidelines.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAXAGE requires the unit to be prompt 1v placed in COLD SHUTDOWN.

3/4.4.8 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential l for Reactor Coolant System leakage or failure due to stress corrosi~on. ,

Maintaining the chemistry within the Steady State Linits ;,rovides adc4uate )

corrosion protection to ensure the structural integril) of the Reactor Coblant i System over the life of the plant. The associated offects of exceeding the oxygen, chloride and fluoride limits are time ad temperature dependent. ,

Corrosion studies .ow that operation may be continued with cont uinant  !

concentration levels in excess of the Steady State limits, up to the Transient  !

Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time l interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.9 SPECIF;C ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

FARLEY - UNIT 2 B 3/4 4-5 AMENDMENT NO.

REACTOR COOLANT SYSTEM BASES The le?kage limit is restricted to 450 GPD (0.315 GPM) total leakage (150 GPD per steam generator) to ensure that offsite doses following a main steam line break will be limited to 10 percent of the 10 CFR 100 guideline.

This restriction is based on the results of a Farley site specific radiological evaluation that assumes a primary coolant iodine activity level corresponding to 1 percent fuel defects (approximately 4.0 microcuries/ gram DOSE EQUIVALENT I-131), rather than a specific activity of 1.0 microcurie / gram DOSE EQUIVALENT I-131, and a post-accident primary-to-secondary leak rate of 55 GPM in the faulted loop.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

l FARLEY - UNIT 2 B 3/4 4-Sa AMENDMENT NO. l I

l l

. f>

Attachment 3 Significant Hazards Consideration Evaluation f

FARLEY NUCLEAR POWER PLANT UNITS 1 AND 2

-TUBE SUPPORT PLATE ELEVATION SG TUBE PLUGGING CRITERION.

SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS INTRODUCTION- e A license amendment is proposed to preclude unnecessarily plugging tubes due to.

the occurrence of outer diameter initiated stress corrosion cracking (00 SCC) at the tube support plate elevations in the Farley Units 1 and 2 steam generators.

Using the existing Technical Specification steam generator tube plugging limit of 40% allowable tube wall penetration, many of the tubes with crack indications would needlessly have to _ be removed from service. - The alternate plugging

-criterion for tube support plate' elevation OD SCC occurring in the Farley Units 1 and 2 steam generators may result in tubes with both partial and through-wall cracks returning to service. In the limiting case, it-is demonstrated that the presence of through-wall cracks alone is not reason enough to remove a tube from service.

DESCRIPTION OF_THE AMEN 0 MENT REQUEST As required by 10 CFR 50.91 (a)(1), an analysis is provided to demonstrate that the proposed: license amendment to implement an alternate steam generator tube plugging criterion for the tube support plate elevations at Farley Units 1 and 2 involves- no significant hazards considerations. The alternate plugging

-criterion involves a correlation between eddy current bobbin probe signal amplitude _ (voltage) and tube burst and leakage capability. The plugging criterion is based on testing of laboratory induced 00 SCC specimens, extensive

examination of pulled tubes from operating steam generators, and field experience f from leakage due to indications at the tube support plates.

Specifically, crack indications with bobbin probe voltages less than or equal to 4.0 volts, regardless of indicated depth, do not require remedial t.ction if

postulated : steam line break leakage can be shown to be acceptable. Crack indications with bobbin- probe signal amplitudes exceeding 4.0 volts, and confirmed to be outer diameter stress corrosion cracking by the rotating pancake coil probe, are plugged or ret ired.

The . proposed amendment would modify Technical Specifications 3/4.4.6 " Steam F Generators," 3/4.4.7 " Reactor Coolant System Leakage," and the associated bases which provide tube. inspection requirements and acceptance criteria to determine the level of degradation for which a tube experiencing OD SCC at the tube support

-plate elevations may remain in service in the Farley Units 1 and 2 steam

-generators.

EVALUATION-i-

L Steam Generator Tube Inteority Discussion l In the development of the alternate plugging criterion, R.G. 1.121, " Bases for Plugging __ Degraded PWR Steam Generator Tubes," and R.G. 1.83, " Inservice o L i

Inspection of PWR Steam Generator Tubes," are used as the bases for determining that steam generator tube integrity considerations are maintained within acceptable limits. R.G.1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria 2,14,15, 31, and 32 by reducing the probability and consequences of steam generator tube rupture through determining the limiting safe conditions of tube wall degradation beyond which tubes with unacceptable cracking, as established by inservice inspection, should be removed from service by plugging. This regulatory guide uses safety factors on loads for tube burst that are consistent with the reouirements of Section 111 of the ASME Code. For the tube support plate elevation degradation occurring in the Farley steam generators, tube burst criteria are inherently satisfied during normal operating conditions by the presence of the tube support plate. The presence of the tube support plate enhances the integrity of the degraded tubes in that region by precluding tube deformation beyond the diameter of the drilled hole. It is not certain whether the tube support plate would function to provide a similar constraining effect during accident condition loadings in either Farley Units 1 or 2. Therefore, no credit is taken in the development of the plugging criterion for the presence of the tube support plate during accident condition loadings.

Conservatively, based on the existing data base, burst testing shows that the safety requirements for tube burst margins during both normal and accident condition loadings can be satisfied with bobbin coil signal amplitudes less than 6.85 volts, regardless of the depth of tube wall penetration of the cracking.

R.G.1.83 describes a method acceptable to the NRC staff for implementing GDC 14, 15, 31, and 32 through periodic inservice inspection for the detection of significant tube wall degradation.

Upon implementation of the plugging criterion, tube leakage considerations must also be addressed. It must be determined that the cracks will not leak excessively during all plant conditions. For the alternate tube plugging criterion developed for the Farley Units 1 and 2 steam generator tubes, no leakage is expected during normal operating conditions even with the presence of through-wall cracks. This is the case as the stress corrosion cracking occurring in the tubes at the support plate elevations in the Farley steam generators are short, tight, axially oriented macrocracks separated by ligaments of material. -

No leakage during normal operating conditions has been observed in the field for crack indications with signal amplitudes less than 7.7 volts. klative to the expected leakage during accident condition loadings, the limiting event with respect to primary-to-secondary leakage is a postulated steam line break (SLB) event. Laboratory data for pulled tubes and model boiler specimens show limited leakage for indications under 10.0 volts during a postulated SLB condition.

Additional Considerations The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging operations. The proposed amendment would minimize the loss of margin in the reactor coolant flow through the steam generator in LOCA analyses. The proposed amendment would avoid loss of margin in reactor coolant system flow and, therefore, assist in demonstrating that minimum flow rates are maintained in excess of that required for operation at full power. Reduction in the amount of tube plugging required can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.

ANALYSIS (3 FACTOR TEST)

In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amendcent is analyzed using the following standards and found not to: 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.

Conformance of the proposed amendment to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:

1) Operation of Farley Units 1 and 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Testing of model boiler specimens for free standing tubes at room temperature conditions show burst pressures as high as 5000 psi for indications of outer diameter stress corrosion cracking with voltage measurements as high as 30 volts. Burst testing performed on pulled tubes with up to 10 volt indications show burst pressures in excess of 5900 psi at room temperature. Correcting for the effects of temperature on material properties and minimum strength levels (as the burst testing was done at room temperature), tube burst capabilit, significantly exceeds the R.G.

1.121 criterion requiring the maintenance of a margin of three times normal operating pressure differential on tube burst if through-wall cracks are present. Based on the existing data base, this criterion is satisfied with bobbin coil indications with signal amplitudes less than 6.85 volts, regardless of the indicated depth measurement. This structural limit is based on a lower 05% confidence level limit of the data. The 4.0 volt plugging criterion compares favorhbly with the structural limit considering expected growth rates of OD SCC at Farley. Alternate crack morphologies can correspond to 6.85 volts so that a unique crack length is not defined by a j burst pressure to voltage correlation. However, relative to expected i leakage during normal cperating conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts (as compared to the 4.0 volt proposed alternate tube plugging limit).

! Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the environment are Loss of External Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe Failure is the most limiting for Far16y Units 1 and 2 in considering the potential for off-site doses. Upon l implementation of the alternate plugging criterion, it must be verified that the expected distribution of cracking indications at the tube support plate intersections are such that primary-to-secondary leakage would result in site boundary doses within a small fraction of the 10 CFR 100 guideline during a postulated steam line break event. Data indicate that a threshold j

l l l

voltage of 2.8 volts would result in through-wall cracks long enough to leak at SLB conditions. A: iication of the proposed plugging criterion requires that the current distribution of number of indications versus voltage be obtained during each refueling outage. The current voltage is then combined with the rate of change in voltage measurement to establish an end of cycle voltage distribution and, thus, leak rate during SLB pressure differential.

If it is found that the potential SLB leakage for degraded intersections planned to be left in service exceeds 55 gpm, then additional tubes will be plugged to reduce SLB leakage potential below 55 gpm.

2) The proposed license amendment does not create the possibility of a new of different kind of accident from any accident previously evaluated.

Implementation of the proposed alternate tube support plate elevation steam generator tube plugging criterion does not introduce any significant changes to the plant design basis. Use of the criterion does not provide a mechanism which could result in an accident outside of the region of the tube support plate elevations. Neither a single or multiple tube rupture event would be expected in a steam generator in which the plugging criterion has been applied (during all plant conditions). The bobbin probe signal amplitude plugging criterion is established such that operational leakage or excessive leakage during a postulated steam line break condition is not anticipated.

Alabama Power Company will implement a maximum leakage rate limit of 150 gpd per steam generator to help preclude the potential for excessive leakage during all plant conditions upon application of the plugging criterion. The R.G.1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture. The 150 gpd limit should provide for leakage detection and plar.t shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible crack length. R.G.1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of three against bursting at normal operating pressure differential. A voltage amplitude of 6.85 volts for typical OD SCC corresponds to meeting this tube burst requirement at the lower 95% uncertainty limit on the burst correlation. Alternate crack morphologies can correspond to 6.85 volts so that a unique crack length is not defined by the burst pressure versus voltage correlation. Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the " longest permissible crack" for evaluating operating leakage limits.

The single through-wall crack lengths that result in tube burst at three i times normal operating pressure differential and SLB conditions are about

! 0.42 inch and 0.84 inch, respectively. Normal leakage for the.se crack lengths would range from 0.11 gpm to 4.5 gpm, respectively, while lower 95%

confidence level leak rates would range from about 0.02 gpm to 0.6 gpm, respectively.

I

- An operating leak rate of 150 gpd will be implemented in application of the 1 tube plugging limit. This leakage limit provides for detection of 0.4 inch t long cracks at nominal leak rates and 0.6 inch long cracks at the lower 95%

confidence level leak rates. Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for SLB conditions at leak rates less than a lower 95% confidence level and for three times normal i' operating pressure differential at less than nominal leak rates.

l The proposed license amendment does not involve a significant reduction in i 3) margin of safety.

{

I The use of the alternate tube support plate elevation plugging criterion at Farley Units 1 and 2 is demonstrated to maintain steam generator tube integrity commensurate with the requirements of Reg. Guide 1.121. R.G.

1.121 describes a method acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by reducing the probability of the consequences of steam l generator tube rupture. This is accomplished by determining the limiting

! conditions of degradation of steam generator tubing, as established by inservice inspection, for which tubes with unacceptable cracking should i,o removed from service. Upon implementation of the criteria, even under the worst case conditions, the occurrence of OD SCC at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions. The most limiting effect would be a possible increase in leakage during a steam line break event.

Excessive leakage during a steam line break event, however, is precluded by verifying th::t, once the criterion is applied, the expected end of cycle 4

distribution of crack indications at the tube support plate elevations would result in minimal, and acceptable primary to secondary leakage during all plant conditions and, hence, help to demonstrate radiological conditions are less than a small fraction of the 10 CFR 100 guideline.

In addressing the combined effects of LOCA + SSE on the steam generator component (as required by GDC 2), it has been determined that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result of lateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings. Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

There are two issues associated with steam generator tube collapse. First, the collapse of steam generator tubing reduces the RCS flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase Peak Clad Temperature (PCT). Second, there is a potential that partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse.

Consequently, Alabama Power Company has performed a detailed leak-before-break analysis and it is concluded that the leak-before-break methodology (as permitted by GDC 4) is applicable to the Farley Units 1 and 2 reactor coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in 6

the structural design basis- of the- plant. : Excluding breaks.-.in the RCS :

primary loops, the _ LOCA loads from the large1 branch line breaks were analyzed at Farley Units l' and 2 and were found- to be of insufficient magnitude- to result in steam' generator tube collapse or significant

. deformation.

Regardless of whether or not leak-before-break is applied to the primary

. loop' piping at Farley. Units 1 and 2, any flow area' reduction is expected to be minimal -(much less than 1%) and PCT margin is available-to account for

. this potential effect. Based-.on recent analyses results, no tubes near wedge-locations are. expected; to- collapse or deform to the degree that

. secondary 'to primary in-leakage would be increased over current expected levels. For all. other steam generator. tubes, the possibility of secondary-

- to-primary leakage in the event of a LOCA +- SSE ' event is not significant.

In actuality, the-amount of secondary-to-primary leakage in the event of a LOCA + SSE is expected to be less than that currently allowed, i.e.,- 500 gpd per steam generator. Furthermore, secondary-to-primary in-leakage would be

. -less than primary to secondary leakage for the same pressure differential

. since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also, the presence of the tube support plate is expected to-reduce the amount of in-leakage.

' Addressing R.G.1.83 considerations, implementation of the tube plugging criterion is supplemented. by 100% inspection requirements at the tube support plate elevations having OD SCC indications, reduced operating leak rate limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating pancake coil inspection requirements for.

the larger indications left in service to characterize the principal degradation mechanism as OD SCC.

As noted previously, implementation of the tube support plate elevation plugging criterion will-decrease-the number of tubes which must be taken out of-service with tube plugs. The installation of steam generator tube plugs

. reduce the RCS flow margin, thus implementation of the alternate plugging criterion will- maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.

Based on the above, it is concluded that the proposed change does not result in a significant reducticn in margin with respect to plant safety as defined in the Final Safety An_alysis Report or any bases of the plant Technical Specifications.

, CONCLUSION Based on the preceding. analysis, it is concluded that using the TSP elevation bobbin coil probe signal amplitude alternate steam generator tube plugging criterion for removing tubes from service at Farley Units 1 and 2 is acceptable and - the proposed license amendment does not involve a Significant Hazards Consideration Finding as defined in 10 CFR 50.92.

1

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  • Attachment 4 Technical Justification
1. WCAP-12871, Revision 1 - J. M. Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates (Proprietary).
2. WCAP-12872, Revision 1 - J. M. Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates (Non-Proprietary).

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