ML20098C185
| ML20098C185 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/25/1995 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20098C180 | List: |
| References | |
| NUDOCS 9510060333 | |
| Download: ML20098C185 (18) | |
Text
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1.
All nonplugged tubes that previously had detectable wall penetrations (>20%).
2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to Specification 4.4.5.4.a.9) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
4.
Indications left in service as a result of application of the expansion transition region circumferential crack repair criteria shall be inspected by plus point or equivalent probe during all future refueling outages.
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 imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found.
d.
Implementation of the expansion transition region circumferential crack repair criteria requires a 100-percent inspection of the hot-leg sludge pile region.
The result of each sample inspection shall be classified into one to the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the 4
inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.
ARKANSAS - UNIT 2 3/4 4-7 Amendment No. MB, s
9510060333 950925 PDR ADOCK 05000368 P
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this Specification 1.
Tubing or Tube means that portion of the tube or sleeve which forms the primary system to secondary system pressure boundary.
2.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
3.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
4.
Degraded Tube means a tube containing imperfections 220% of nominal wall thickness caused by degradation.
5.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
6.
Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube containing a defect is defective.
7.
Pluqqing or Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next inspection. The plugging or repair limit is equal to 40% of the nominal parent tube and sleeve wall thickness for sleeves installed in accordance with B&W Topical Report BAW-2045-PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design -
Application to ANO Unit 2".
The plugging limit is equal to 34%
of the nominal sleeve wall thickness for sleeves installed in accordance with CENS Report CEN-601-P, "ANO-2 Steam Generator l
Tube Repair Using Leak Tight Sleeves", Revision 01-P, dated July, 1992.
This definition does not apply to indications for which the expansion transition region circumferential crack repair criteria are being applied.
Refer to 4.4.5.4.a.ll for the repair limit applicable to these indications.
8.
Unserviceable describes the condition of a tube 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 i
break as specified in 4.4.5.3.c, above.
9.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely i
around the U-bend to the top support of the cold leg.
ARKANSAS - UNIT 2 3/4 4-9 Amendment No. 434,44G, 1
SURVEILLANCE REQUIREMENTS (Continued) l 10.
Preservice 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 f
OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11.
Expansion Transition Region Circumferential Crack Plugging or Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing circumferentially oriented outside diameter stress corrosion cracking within the expansion transition region at the top of the tubesheet. Within the expansion transition region, the plugging or repair limit is based on maintaining steam generator tube serviceability as i
described below:
a.
Steam generator tubes, whose degradation is attributed to circumferentially oriented outside diameter stress corrosion cracking within the expansion transition area and have a degraded area less than 40% of the tube wall cross sectional area, will be allowed to remain in service, b.
Steam generator tubes, whose degradation is attributed to circumferentially oriented outside diameter stress corrosion cracking within the expansion transition area and have a degraded area equal to or greater than 40% of the tube wall cross sectienal area, will be repaired or plugged, b.
The steam generator shtall be determined OPERABLE after completing the corresponding acticns (plug or repair all tubes exceeding the plugging or repair limits) required by Table 4.4-2.
Defective tubes l
may be repaired in accordance with:
1)
B&W Topical Report BAW-2045PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design-Application to ANO Unit 2".
2)
CENS Report CEN-601-P, "ANO-2 Steam Generator Tube Repair Using Leak Tight Sleeves", Revision 01-P, dated July, 1992.
4.4.5.5 Reports a.
Following each inservice inspection of steam generator tubes the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.
b.
The complete results of the steam generator tube inservice inspection shall be reported on an annual basis for the period in which the inspection was completed. This report shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged or sleeved.
I ARKANSAS - UNIT 2 3/4 4-10 Amendment No. -143, 444,
REACTOR CCOLANT SYOTEM t
SURVEILLANCE REQUIREMENTS (Continued) c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report pursuant to Specification 6.9.2 as denoted by Table 4.4-2.
Notification of the Commission will be made prior to resumption of plant operation. The written Special Report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
For implementation of the expansion transition region circumferential
- d..
crack repair criteria, notify the staff prior to returning the steam generators to service should any of the following conditions arise:
1.
If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) circumferential crack size distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating' cycle.
2.
If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) circumferential crack size distribution exceeds 1 X 10'#, notify the NRC and provide an assessment of the safety significance of the occurrence.
3.
If indications are identified within the expansion transition region that are attributable to primary water stress corrosion cracking.
l j
L ARKANSAS - UNIT 2 3/4 4-10a Amendment No. 91,
f REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE I
LIMITING CONDITION FOR OPERATION Reactor Coolant System leakage shall be limited to 3.4.6.2 No PRESSURE BOUNDARY LEAKAGE, a.
b.
1 GPM UNIDENTIFIED LEAKAGE, 150 gallons per day of primary-to-secondary leakage through any one c.
steam generator, and 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, d.
Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6.1.
e.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
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 a.
hours.
than any one of With any Reactor Coolant System leakage greatet reduce the b.
excluding PRESSURE BOUNDARY LEAKAGE, the above limits, 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 />.
With any Reactor Coolant System Pressure Isolation Valve leakage isolate the high pressure portion c.
greater than the above limit,of the affected system from the low pressure portion hours by use of at least two valves
- in each high pressure line having a non-functional valve and 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 />.
- These valves may include check valves for which the leakage rate has Manual and automatic been verified, manual valves or automatic valves.
valves shall be tagged as closed to preclude inadvertent valve opening.
ARKANSAS - UNIT 2 3/4 4-14 Amendment No.
Order d:t
'/20/91
l i
REACTOR COOLANT SYSTEM BASES Demonstration of the safety valves' lift setting will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves against water relief. The steam bubble functions to relieve RCS pressure during all design transients.
The requirement that 150 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss-of-offsite power condition to maintain natural circulation at HOT STANDBY.
3/4.4.5 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 naintained.
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 ta 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 (primary-to-secondary leakage = 150 gallons per day through any one steam generator).
Cracks l 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.
Primary-to-secondary leakage of 150 gallons per day through any one steam generator can readily be detected by steam generator blowdown radiation monitors or N-16 monitors.
Leakage in excess of this limit will require plant shutdown and an unscheduled 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,
.c will be found during scheduled inservice steam generator tubes examinLtions.
Plugging or sleeving will be required for all tubes with imperfections exceeding the plugging or repair limit as defined in Surveillance Requirement 4.4.5.4.a.
Defective tubes may be repaired by sleeving in accordance with the B&W Topical Report BAW-2045PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design-Application to ANO Unit 2" or CENS Report CEN-601-P, "ANO-2 Steam Generator Tube Repair Using Leak Tight sleeves",
l Revision 01-P, dated July 1992.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the tube wall thickness.
For sleeved tubes, the adequacy of i
ARKANSAS - UNIT 2 B 3/4 4-2 Amendment No. GO,-143,44G,MB, 1
REACTOR COOLANT SYSTEM BASES I
the system that is used for periodic inservice inspection will be validated.
Additionally, upgraded testing methods will be evaluated and appropriately implemented as better methods are developed and validated for commercial use.
The circumferential crack repair limit, which utilizes the percentage of degraded area of the tube wall thickness as opposed to a traditional maximum depth, is applicable only for circumferentially oriented indications caused by outside diameter stress corrosion cracking in the expansion transition area at the top of the tubesheet. All circumferential indications previously identified in the ANO-2 steam generators have occurred in the hot-leg sludge pile region.
The sludge pile region is bounded by an arc between lines 10 and 156 extending up to row 122 of a steam generator.
1 The circumferential crack repair limit of 40% was determined by performing a structural analysis per the recommendations of draft Regulatory Guide 1.121 and applying the following uncertainties: 959, lower bound material properties, 95%
lower bound burst curve, 95% lower bound eddy current measurement uncertainties, and 95% upper bound crack growth rate.
Specification 4.4.5.5.d implements several reporting requirements for which the NRC will be notified prior to returning the steam generators to service.
For the purpose of this reporting requirement, leakage and conditional burst probability can be calculated on the as-found circumferential crack distribution rather than the projected end-of-cycle circumferential crack distribution when it is not practical to complete these calculations using the projected end-of cycle distribution prior to returning the steam generators to service. Note that if leakage and conditional burst probability were calculated using the as-found circumferential crack distribution for the purposes of addressing the Specification 4.4.5.5.d reporting criteria, then the results of the projected end-of-cycle distribution will be provided within 90 days following restart after a steam generator inspection.
If it is determined following restart that the end of cycle conditional burst probability or leakage would exceed the applicable limits, the technical specification reactor coolant system iodine activity limits may be lowered or the reactor operating cycle shortened in order to remain within the applicable limits.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3 certain results will be reported in a Special Report to the Commission pursuant to Specification 6.9.2 as denoted by Table 4.2-2.
Notification of the Conadssion will be made 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 necessary.
3/4. 4. 6 REACTOR COOLANT SYSTEM LEAYAGE 3/4.4.6.1 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.
l ARKANSAS - UNIT 2 B 3/4 4-3 Amendment No. M,433,
REACTOR C00LAMT SYSTEM BASES 3/4.4.6.2 REACTOR COOLANT SYSTEM 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 allowances 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 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 as a portion of the allowed limit.
The leakage limit of 150 gallons per day for each steam generator ensures l
that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. This leakage limit is more restrictive than the standard i
operating leakage limit and is intended to provide an additional margin to j
accommodate a circumferential crack which might grow at a greater than expected 1
rate.
Hence, the reduced leakage limit, when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will be detected, and the plant shut down in a timely manner.
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.
)
3/4.4.7 CHEMISTRY l
The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce 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.
3/4.4.8 SPECIFIC ACTIVITY The lindtations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a
)
l ARKANSAS - UNIT 2 B 3/4 4-4 Amendment No.
Order-dated i 20 01
E 4
MARKUP OF CURRENT ANO-2 TECHNICAL SPECIFICATIONS (FORINFORMATION ONLY) 4
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1.
All nonplugged tubes that previously had detectable wall penetrations (>20%).
2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to Specification 4.4.5.4.a.9) shall be performed on each selected tube.
If any selected tube does not perndt the passage of the eddy current probe for a tube inspection, this shall-be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
1 l
4.
Indications left in service as a result of application of the j
expansion transition recion circumferential crack repair criteria
]
shall be inspected by plus coint or equivalent probe durina all i
future refuelina outaces.
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 imperfections were previously found.
2.
The' inspections include those portions of the tubes where l
. imperfections were previously found.
d.
Implementation of the expansion transition recion circumferential crack repair criteria reauires a 100-oercent inspection of the hot-lea sludae oile recion.
The result of each sample inspection shall be classified into one to the following three categories:
J i
Category Inspection Results C-1 Less than 5% of the total tubes inspected l
are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than l% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
I C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the l
4 inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.
l 1
. ARKANSAS - UNIT 2 3/4 4-7 Amendment No. 168,
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance criteria a.
As used in this Specification 1.
Tubing or Tube means that portion of the tube or sleeve which forms the primary system to secondary system pressure boundary.
2.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
3.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
4.
Degraded Tube means a tube containing imperfections 220% of nominal wall thickness caused by degradation.
5.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
6.
Defect means an imperfection of such severity that it exceeds the plugging ot repair limit. A tube containing a defect is defective.
7.
Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next inspection. The plugging or repair limit is equal to 40% of the nominal parent tube and sleeve wall thickness for sleeves installed in accordance with B&W Topical Report BAW-2045-PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design -
Application to ANO unit 2".
The plugging limit is equal to 34%
of the nominal sleeve wall thickness for sleeves installed in accordance with CENS Report CEN-601-P, "ANO-2 Steam Generator Tube Repair Using Leak Tight Sleeves", Revision 01-P, dated July, 1992.
This definition does not apply to indications for which the expansion transition recien circumferential cra_c_k repair criteria are beina applied.
Refer to 4.4.5.4.a.ll for the repair limit applicable to these indications.
8.
Unserviceable describes the condition of a tube 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 4.4.5.3.c, above.
9.
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.
4 i
ARKANSAS - UNIT 2 3/4 4-9 Amendment No. 434,444, l
i
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 10.
Preservice 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.
Expansion Transition Recion Circumferential Crack Pluccino or Pepair Limit is used for the disposition of a steam cenerator tube for continued service that is experiencino circumferentially oriented outside diameter stress corrosion crackinc within the expansion transition recion at the top of the tubesheet. Within the expansion transition recion, the pluccino or repair limit is based on maintaininc steam cenerator tube serviceability as described below a.
Steam cenerator tubes, whose decradation is attributed to circumferentially oriented outside diameter stress corrosion crackinc within the expansion transition area and have a decraded area less than 40% of the tube wall cross sectional area, will be allowed to remain in service.
b.
Steam _ generator tubes, whose decradation is attributed to circumferentiallv_ oriented outside diameter stress corrosion crackinc within the expansion transition area and have a
+
decraded area ecual to or creater than 40% of the tube wall cross sectional area, will be repaired or plucced.
j b.
The steam generator shall be determined OPERABLE after completing i
the corresponding actions (plug or repair all tubes exceeding the plugging or repair limits-end all tube: centaining through-aall eeeeke) required by Table 4.4-2.
Defective tubes may be repaired in accordance with:
1)
B&W Topical Report BAW-2045PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design-Application to ANO Unit 2".
2)
CENS Report CEN-601-P, "ANO-2 Steam Generator Tube Repair Using Leak Tight sleeves", Revision 01-P, dated July, 1992.
4.4.5.5 Reports a.
Following each inservice inspection of steam generator tubes the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.
b.
The complete results of the steam generator tube inservice inspection shall be reported on an annual basis for the period in which the inspection was completed. This report shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged or sleeved.
l ARKANSAS - UNIT 2 3/4 4-10 Amendment No. 4M, 44,
REACTOR COCLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in_a Special Report pursuant to Specification 6.9.2 as denoted by Table 4.4-2.
Notification of the Commission will be made prior to resumption of plant operation. The written Special Report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
d.
For implementation of the expansion transition recion circumferential crack repair criteria, notify the staff orior to returnina the steam cenerators to service should any of the followino conditions arisg1 1.
If estimated leakaae based on the croiected end-of-evele (or if not oractical, usina the actual measured end-of-evele) circumferential crack size distribution exceeds the leak limit (determined from the licensina basis dose calculation for the costulated main steam line break) for the next operatina cycle.
2.
If the calculated conditional burst probability based on the croiected end-of-cycle (or if not practical, usina the actual measured end-of-cyclel circumferential crack size distribution exceeds 1 X 10", notify the NRC and provide an assessment of the safety sianificance of the occurrence.
3.
If indications are identified within the exoansion transition reaion that are attributable to primary water stress corrosion crackina.
ARKANSAS - UNIT 2 3/4 4-10a Amendment No. M,
REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, c.
1 CP" total pri= ry t: ::::ndary 10 hage through b;th st;;r g ncreter and 0.5 CP" th ugh :ny en :tcar generater, 150 callons per das of primarv-to-secondary leakace throuah any one steam cenerator, d.
10 GPM IDr.NTIFIED LEAKAGE from the Reactor Coolant System, and Leakage as specified in Table 3.4.6-1 for those Reactor Coolant e.
System Pressure Isolation Valves identified in Table 3.4.6.1.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY a.
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
- hours, b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, With any Reactor Coolant System Pressure Irolation Valve leakage c.
greater than the above limit, 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 valves
- in each high pressure line having a non-functional valve and 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 />.
- These valves may include check valves for which the leakage rate has been verified, manual valves or automatic valves.
Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.
ARKANSAS - UNIT 2 3/4 4-14 Amendment No.
Ord:: d:t tl20!?1
)
REACTOR COOLANT SYSTEM BASES Demonstration of the safety valves' _ift setting will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves against water relief. The steam bubble functions to relieve RCS pressure during all design transients.
The requirement that 150 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss-of-offsite power condition to maintain natural circulation at HOT STANDBY.
3/4.4.5 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 naintained. 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 cracxing.
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 (primary-to-secondary leakage = 0.5 CP" 150 callons per day throuah any one pee-steam l
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 i
I imposed during normal operation and by postulated accidents. Operating plant:
have d : nctrated th t-pPrimary-to-secondary leakage of 0.5 CP" 150 callons per day throuch any one pee-steam generator can readily be detected by speam generator blowdown radiation monitors or N-16 monitors.
Of tear gent:stor blewdcun Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking
, 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 tubes examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the plugging or repair limit as defined in Surveillance Requirement 4.4.5.4.a.
Defective tubes may be repaired by sleeving in accordance with the B&W Topical Report BAW-2045PA-00 as supplemented by the information provided in B&W Report 51-1212539-00, "BWNS Kinetic Sleeve Design-Application to ANO Unit 2" or CENS Report CEN-601-P, "ANO-2 Steam Generator Tube Repair Using Leak Tight Sleeves",
Revision 01-P, dated July 1992.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradetion that has penetrated 20% of the tube wall thickness.
For sleeved tubes, the adequacy of ARKANSAS - UNIT 2 B 3/4 4-2 Amendment No. 40,M3,443,MB, i
i s
l I
i BASES l
the system that is used fc periodic inservice inspection will be validated.
Additionally, upgraded testing methods will be evaluated and appropriately implemented as better methods are developed and validated for commercial use.
I The circumferential crack repair limit, which utilizes the percentaae of decraded area of the tube wall thickness as opposed to a traditional maximum depth, is applicable only for circumferential1v oriented indications caused by outside diameter stress corrosion crackino in the expansion transition area at the top of the tubesheet. All circumferential indications previous 1v identified in the_ANO-2 steam cenerators have occurred in the hot-lea sludae pile recion.
The sludge pile recion is bounded by an are between lines 10 and 156 extendina up to row 122 of a steam cenerator.
4 The circumferential crack repair limit of 40% was determined by performino a q
structural analysis per the recommendations of draft Reculatory Guide 1.121 and app! vino the followino uncertainties: 95% lower bound material properties, 95%
lower bound burst curve, 95% lower bound eddy current measurement uncertainties, and 95% upper bound crack crowth rate.
Specification 4.4.5.5.d implements several reportino recuirements for which the NRC will be notified prior to returnino the steam cenerators to service.
For the purpose of this reportinq requirement, leakace and conditional burst probability can be calculated on the as-found circumferential crack distribution
.rather than the proiected end-of-cycle circumferential crack distribution when it is not practical to complete these calculations usino the proiected end-of cycle distribution prior to returnina the steam cenerators to service.
Note that if leakace and conditional burst probability were calculated usina the as-found circumferential crack distribution for the purposes of addressina the Specification 4.4.5.5.d reportino criteria, then the results of the proiected end-of-cycle distribution will be provided within 90 days followina restart after a steam cenerator inspection.
If it is determined followina restart that the end of cycle conditional burst probability or leakace would exceed the applicable limits, the technical specification reactor coolant system iodine activity limits may be lowered or the reactor operatina cycle shortened in order to remain within the applicable limits.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3 certain results will be reported in a Special Report to the Commission pursuant to Specification 6.9.2 as denoted by Table 4.2-2.
Notification of the Commission will be made 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 necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 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.
I ARKANSAS - UNIT 2 B 3/4 4-3 Amendment No. M, H3, m
REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited ar.ount 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 allowances 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 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 as a portion of the allowed limit.
The bebek-eteer generator tub: leakage limit of 1 CP" for all 150 callons per day for each steam generatore ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 CPM limit is eeneletent akth the cacumption: used in th analysi: of th::: :: idents.
The 0.5 GP" 1:chage-14mit--per steer gencr ter eneurc: that stecm gencreter tube integrity le-ma4etained 1.-
the event of a mair sten line rupture :: under LOCA ::adition:.
This leakace limit is more restrictive than the standard operatino leakace limit and is intended to provide an additional margin to accommodate a circumferential crack which micht crow at a creater than expected rate.
Hence, the reduced leakace limit, when combined with an effective leak rate monitorino procram, provides additional assurance that should a sionificant leak be experienced in service, it will be detected, and the plant shut down in a timely manner.
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.
3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce 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 operaticn 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 j
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.
3/4.4.8 SPECIFIC ACTIVITY ARKANSAS - UNIT 2 B 3/4 4-4 Amendment No.
Geder-deted i 20 91
(
The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a l
d ARKANSAS - UNIT 2 B 3/4 4-4 Amendment No.
Order dated '
20 "1