ML17333A477
| ML17333A477 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/19/1996 |
| From: | AMERICAN ELECTRIC POWER CO., INC. |
| To: | |
| Shared Package | |
| ML17333A478 | List: |
| References | |
| AEP:NRC:1166AA, GL-95-05, GL-95-5, NUDOCS 9606260256 | |
| Download: ML17333A477 (45) | |
Text
ATTACHMENT 2 TO AEP:NRC:1166AA EXISTING TECHNICAL SPECIFICATION PAGES MARKED TO REFLECT PROPOSED CHANGES 9606260256 MObi9 POR
- GGCK 050003i5 P
3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4 4 REACTOR COOLANTSYSTEM STEAM GENERATORS LIMITINGCONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION With one or more steam genezatozs inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing T,~ above 200'F.
SURVEILLANCERE UIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented insezvice inspection program and the requirement of Specification 4.0.5.
4.4.5.1 Steam Generator S le 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.
4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4A-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
When applying the exceptions of 4.4.5.2.a through 4AD.2.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection.
The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% ofthe tubes inspected shall be from these critical areas.
The fiat sample of tubes selected for each inservice inspection (subsequent to the pzesezvice inspection) of each steam generator shall include:
l.
AH tubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged or repaired by sleeving in the affected area.
This Specification does not apply in Mode 4 while performing crevice flushing as long as Limiting Conditions for Operation for Specification 3.4.1.3 are maintained.
COOK NUCLEAR &LANT-UNITI Page 3/4 4-7 AMENDMENTQQ, 446, %0, 205
3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLAiVZSYSTEII SURVEILLANCERE UIREMENTS (conunued) 2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to Specitication 4.4.5.4.a.8) shall be performed on each selected tube. Ifany 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.
Tubes left in service as a result of application ofthe tube support plate plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.
c.
In addition to the sample required in 4.4.5.2.b.
1 through 3, all tubes which have had the F~ criteria applied wBI be inspected in the roll expanded region.
The roll expanded region of these tubes may be excluded from the requirements of 4.4.5.2.b.l.
d.
The tubes selected as the second and third samples (ifrequired by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
e.
2.
The inspections include those portions of the tubes where imperfections were previously found.
Implementation of the steam generator tube/tube support plate plugging criteria~te.
support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion crachng (ODSCC) indications.
The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.
f.
Inspection of sleeves wiB follow the initial sample selection (V'ample) and sample expansion requirements of Table 4.4-2.
The results of each sample inspection shall be classified into one of the following three categories:
~Cate orv Ins ection Results C-1 Less than 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 inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% ofthe total tubes inspected are degraded tubes or more than 1%
of the inspected tubes are defective.
COOK NUCLEARPLANT-UNIT 1 Page 3/4 44 AMENDMENT484, 466, KV, 478; %0, 205
3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continuedi 4.4.5.3 Note:
In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.
Ins tion Fre uencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
ae 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 followingservice under AVTconditions, not including the preservice inspection, result in all inspection results falling into the C-l 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.
If the results of 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 4.4.5.3.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 followingconditions:
1.
Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.
2.
A seismic occurrence greater than the Operating Basis Earthquake.
3.
A losswf-coolant accident requiring actuation of the engineered safeguards.
4.
A main steam line or feedwater line break.
COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-9 AMENDMENT98, IGG, KO, 205
3/4 LIMITINGCONDITIOifS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLAiVfSYSTEM SURUEILLAHCERE UIREMENTS (continued) 4.4.5.4 Acce tance Criteria As used in this Specification:
Imoerfection means an exception to the dimensions, finish or coatour of a tube or sleeve 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.
2.
Degradation means a service-induced
- cracking, wastage, wear or general corrosioa occumng on either inside or outside of a tube or sleeve.
3.
De ded Tube or Sleeve meaas an imperfection greater than or equal to 20% of the nominal wall thickness caused by degradation.
Percent De radation means the amount of the tube wall thickness affected or removed by degradation.
Defect means an imperfection of such severity that it exceeds the repair limit.
6.
Repair/Plu in Limit meaas the imperfection depth at or beyond which the tube or sleeved tube shall'be repaired or removed from service.
Any tube which, upon inspection, exhibits tube wall degradatioa of40 percent or more ofthe nominal tube wall thickaess shall be plugged or repaired prior to returning the steam generator to service.
This definition does not apply to the portion of the tube in the tubesheet below the F~
distaace for F~ tubes.
Any sleeve, except laser welded sleeves, which upon inspection exhibits wall degradation of 29 percent or more of the nominal wall thickaess, shall be plugged prior to returning the steam generator to service. In addition, any sleeve, except laser welded sleeves, exhibiting any measurable waII loss in sleeve expansion transition or weld zones shall be plugged.
This definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied.
Refer to 4.4.5.4.a.10 for the plugging limitapplicable to these intersections.
For a tube that has beea sleeved with a laser welded sleeve, thmugh wali penetration ofgreater than or equal to 23% of sleeve nominal wall thickness requires the tube to be removed fmm service by plugging.
7.
Unserviceable describes the condition of a tube or sleeve ifit leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a losswfwoolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.
8.
Innxction determines the conditioa of the steam generator tube or sleeve fmm the point of catty (hot leg side) completely around the U-bend to the top support of the cold leg.
For a tube in which the tube support plate clcvation interim plugging limit has been
- applied, the inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of thc last crack indication.
COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-10 AMENDMENT 98, 4', 466,
~, AO, ZOS
3/4 LIMITINGCONDITIONS FOR OPERATION Ag~ SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANTSYSTEM SURVEILLANCERE UIREMENTS (continued)
~glee>>in a tube is pelmlued with nlbe support plate sleeves and with mbmheet sleeves.
Tube support plate sleeves are centered about the tube support plate intersection.
Tubeshcct sleeves start at the primary Quid tubeshcet face and extend to the free span region of rube above the tubeshect.
10.
ube Sup rt Plate Re air Limit is used for the disposition of a steam generator tube for co 'ed service that is experiencing outside diameter stress corrosion cracking (ODS confined within the thiclmess of the tube support plates. Attube support plate iaterscctio, the repair limitis based on maintaining steam generator tube serviceability as described low:
a.
Degradati attributed to ODSCC within the hounds of thc tube support plate with bobbin tage less than or equal to 2.0 volts willbe allowed to remain in service.
b.
Degradation attributed.
ODSCC within the bounds of the tube support plate with a bobbin voltage grea than 2.0 volts willbe repaired or plugged except
~ as aoted ia 4...4.a.10.c bel ladica ns of potential degradation at buted to ODSCC within the bounds of th be support plate with a bobbin volta reater than 2.0 volts but less than r equal to 5.6 volts may remain in service i tating pancake coil inspectioa does aot detect degradation.
Indications of ODS degradation with a bobbin voltage greater than 6.6 volts will be plug ed or tepahqd.
11.
F'istance is the distance from the bottom of the hardroll transition toward the bottom of the tubeshect that has been conservatively determined to be 1.11 inches (aot including eddy curreat uncertainty).
(2.
F* Tube isa tube with degradation, below the F*distance, equal to orgreater than40%,
aud not degraded (i.e., no indications of cracking) within the F>> distance.
13.
Tube Rcoair refers to sleeving as described by the reports listed in 4.4.5.4.c which are used to maintain a tube in service or return a tube to service.
Tubes with degradation indications of less than the plugging limit may be prcveatively sleeved at the Owner's discretion... This includes removal of plugs that werc installed as a corrective or
~
preventive measure.
A tube inspecuoa per 4.4.5.4.a>>8 is required prior to returning previously plugged tubes to service.
Further restrictioas regarding identified indications and their proximity to the joint areas of various sleeving processes may be applicable.
The steam generator shall be determined OPERABLE after completing the correspoading actions (plugging or slccving all tubes exceeding the'repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.
Stcam generator tube repairs may be made in accordance with thc methods described in either WCAP-12623, %CAP-13088 (Rev. 3), or CEN-313-P.
COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-11 AMENDMENT48k 446 4VV
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10.
Tube Suonort Plate Plu crina Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates.
At tube support plate intersections, the plugging (repair} limit is based on maintaining steam generator tube serviceability as described below:
a ~
- tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.
- tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a
bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 4.4.5.4.a.10.c below.
c ~
d.
- tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.
- tubes, with indications of outside diameter stress corrosion cracking degradation with a
bobbin voltage greater than the upper voltage repair limit'illbe plugged or repaired.
Xf an unscheduled mid-cycle inspection is performed, the foll@wing mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.
(1)
The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05.
t I
The mid-cycle the following repair limits are determined from equations:
Vsc 10+ NDE -'
CL i CL-tr V~= Verve (Vmc Vau.)
CL where:
Vsr.
Gr NDE upper voltage repair limit lower voltage repair limit mid-cycle upper voltage repair limit based on time into cycle mid-cycle lower voltage repair 1'mit based on V~ and time into cycle length of time since last scheduled inspection during which V~ and V~ were implemented cycle length (the time between two scheduled steam generator inspections) structural limit voltage average growth rate per cycle length 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e.,
a value of 20-percent has been approved by NRC)
Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.
C.
(2)
The NDE is the value provided by the NRC in GL 95-05-~~
LIMITINGCOND1TIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 4.4.5.5
~Re otto Following each inservice inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.
The complete results of the steam generator mbe inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.
This report shall include:
1.
Number and extent of tubes and sleeves inspected.
2.
Location and percent ofwalI-thickness penetration for each indication of an imperfection, 3.
Identification of tubes plugged or sleeved.
C.
Results of steam generator rube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation.
The wrinen foBowup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
d.
F
'lementation of the voltage-based repair criteria to tube support plate interactions, notify the s rior to retuaiing the steam generators to service should any of the following conditions arise:
Ifestimated 1
ge based on the actual measured endwf~cle voltage distribution would have exceeded the limit(for the postulated main steam line break utilizing Standard Review Plan - N 800" assumptions) during the previous operating cycle.
If circumferential crack-like 'ications are detected at the tube support plate intersections.
Ifsignificant indica are identified that e end beyond the confines ofthe tube support plate.
If calculated conditional burst probability, as calcu ed per WCAP-14277, exceeds x10', notify the NRC and provide an assessment of the, safety significance of the occurrence.
COOK NUCLEAR PLANT-UNITI Page 3/4 4-12 AMENDMENT96, 466, %VV, 478, 200, 205
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ZivQev P d.
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission 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) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.
2 4 3 ~
If circumferential crack-like indications are detected at the tube support plate intersections.
If indications are identified that extend beyond the confines of the tube support plate.
4
~
If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1
X 10
, notify the Commission and, provide an assessment of the 'safety significance of the occurrence.
3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANTSYSTEM OPERATIONAL LEAKAGE LIMITINGCONDITION FOR OPERATION 3 4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARYLEAKAGE, b.
1 GPM UNIDENTIFIEDLEAICAGE, C.
e.
600 gallons per day total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generato~
~du 10 GPM IDENTIFIEDLEAKAGEfrom the Reactor Coolant System, Seal line resistance greater than or equal to 2.27E-1 ft/gpm'nd, f.
The leakage from each Reactor Coolant System Pressure Isolation Valves specified in Table 3.44 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System average pressure within 20 psi of the nominal full pressure value.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With any PRESSURE BOUNDARYLEAKAGE, be in at least HOT STANDBYwithin 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 BOUNDARYLEAKAGE,reduce the leakage rate to within limits within4 hours or be in at least HOT STANDBYwithin the nexx 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(s) leakage greater than the above limit, declare the leaking valve inoperable and isolate the high pressure portion of the affected system from the low pressure portion by the use of a combination of at least two closed valves, one of which may be the OPERABLE check valve and the other a dosed de-energized motor operated valve. Verifythe isolated condition ofthe dosed de~ergized motor operated valve at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-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 />.
SpeciIIcation 3.4.6.2.e is applicable with average pressure within20 psi ofthe nominal fullpressure value.
COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-16 AMENDMENT442~ 446 478, 488t 200
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3/4 BASES 3/4.4 REACTOR COOLANTSYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY The Surveillance Requirements for inspection ofthe steam generator tubes ensure that the structural integrity ofthis portion of the RCS willbe 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 insoection 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 w' be maintained within those chemistry lunits found to result in negligible corrosion of the steam generator chemisuy is not maintained within these parameter limits, lo&ized corrosion crachng.
The extent ofcracking during plant operation would be limited by the leakage between the primary coolant system and the secondary coolant system. Th bes.
Ifthe secondary coolant likely result in stress corrosion tation ofsteam generator tube allowable primaty-to-secondaty 15).
Axial or circumferentially oriemed cracks having a pritnaty-to-secondaty leakage less than this limitduring operation willhave an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Leakage in excess of this limit willrequire plant shutdown and an inspection, during which the leaking tubes willbe located and plugged or repaired.
A steam generator while undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/ operating upon reinstatement ofauxiliary or main feed flowcontrol and steam control.
Wastage-type defects are unlikely with the all volatile treaunent (AVT)of secondary coolant.
However, even if a defect of similar type should develop in service, it willbe found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the repair limit
. which is defined in Specification 4.4.5A.a. Steam generator tube inspections ofoperating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wali thickness.
Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criteria of4.4.5.4.a.10.
COOK NUCLEAR PLANT-UNITI Page B 3/4 4-2n AMENDMENTQQ; kR, 466, 478, 200
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iv58'-4 Q.
The voltage-based repair limits of these surveillance requirements (SR) implement the guidance in GL 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections.
The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at. other locations within the SG.
Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no NDE detectable cracks extending outside the thickness of the support plate.
Refer to 6L 95-05 for additional description of the degradation morphology.
Implementation of these SRs requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure/bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650'F (i.e.,
the 95-percent LTL curve).
The voltage structural limit must be adjusted downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit; V~,
is determined from the structural voltage limit by applying the following equation:
V~ = VSL - VQz VNDE where Vq~ represents the allowance for degradation growth between inspections and V~q represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
The mid-cycle equation in SR 4.4.5.4.10.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning the SGs to service.
For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle (EOC) voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service.
Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.l and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per the GL section 6.b (c) criteria.
3/4 BASES 3/4.4 REACTOR COOLANTSYSTE>I 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.
3/4.4.6.2 OPERATIONALLEAKAGE 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 IDENTIFIEDLEAXAGElimitations provides allowance for a limited amount ofleakage from known sources whose presence will not interfere, with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The limitation on seal line resistance ensures that the seal line resistance is greater than or equal to the resistance assumed in the minimum safeguards LOCAanalysis.
This analysis assumes that all of the flow that is diverted from the boron injection line to the seal injection line is unavailable for core cooling.
$ g minimize the potential for a large leakage even during steam e break under LO A conditions.
Based on the NDE uncertainties, bobbin coil voltage distributi n and crack gr wth rate from the p ious inspection, the expected leak rate following a steam line rupture is limite to below~gpm which will the calculated offsite doses ~4
~~4 within 10 percent of the 10 CFR 100 Leakage in the intact loops is limited to 150 gpd. Ifthe projected end of cycle distribution of mdications results in primary-to-secondary leakage greater than gpm in the faulted loop during a postul steam line break event, additional tubes must be removed from servi in order to reduce the postulated p 'to-secondary steam line break leakage to below m.
PRESSURE BOUNDARY GE ofany magnitude is unacceptable since it may be indicative ofan impending gross failure of the pressure undary.
Should PRESSURE BOUNDARYLEAKAGEoccur through a component which can be isolated fro e balance of the Reactor Coolant System, plant operation may continue provided the leaking component is p mptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.
Ojj COOK NUCLEAR PLANT-UMT1 Page B 3/4 4-3 AMENDMENT%,A,4'; ~, 200
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Also, the 150 gallons per day leakage limit incorporated into this specification is more restrictive than the standard operating leakage limit and is intended to provide an additional margin to accommodate a crack which might grow at a
greater than expected rate or unexpectedly extend outside the thickness of the tube support plate.
- 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.
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ATTACHMENT 3 TO AEP:NRC:1166AA r
PROPOSED REVISED TECHNICAL SPECIFICATION PAGES
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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 2.
Tubes in those areas where experience has'indicated potential problems.
3.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. Ifany 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.
Tubes left in service as a result of application of the tube support plate plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.
c.
In addition to the sample required in 4.4.5.2.b.1 through 3, all tubes which have had the F~ criteria applied will be inspected in the roll expanded region.
The roll expanded region of these tubes may be excluded from the requirements of 4.4.5.2.b.l.
d.
The tubes selected as the second and third samples (ifrequired by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
The inspections include those portions of the tubes where imperfections were previously found.
Implementation of the steam generator tube/tube support plate plugging criteria requires a 100 percent bobbin coil inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.
The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.
Inspection of sleeves will follow the initial sample selection (1s'ample) and sample expansion requirements of Table 4.4-2.
The results of each sample inspection shall be classified into one of the following three categories:
~Cate ~o Ins ection Results C-1 Less than 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 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.
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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 9.
~Sleavin a tube is permiued with tube support plate sleeves aud with tubesheet sleeves.
Tube support plate sleeves are centered about the tube support plate intersection.
Tubesheet sleeves start at the primary fluid tubesheet face and extend to the free span region of tube above the tubesheet.
10.
Tube Su ort Plate Re air Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates.
At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:
Steam generator tubes, whose degradation is attributed to ouside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts willbe allowed to remain in service.
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 4.4.5.4.a. 10.c below.
C.
Steam generator tubes with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds ofthe tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limitI may remain in service ifa rotating pancake coil or acceptable alternative inspection does not detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged or repaired.
Ifan unscheduled mid-cycle inspection is performed, the following midwycle repair limits apply instead ofthe limits identified in 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.
The upper voltage repair limit is calculated according to the methodology in Generic Letter 9545.
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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued)
The mid-cycle repair limits are determined from the following equations:
Vs>,
1.0 + NDE + Gr CL-~t Vuzm = Vent,
( Vivat, Vivat, )
CL where:
VURL
=
upper voltage repair limit VLRL =
lower voltage repair limit VMURL =
mid-cycle upper voltage repair limit based on time into cycle VMLRL=
mid-cycle lower voltage repair limit bassed on VMIJRLand time into cycle ht
=
Length of time since last scheduled inspection during which VURL and VLRLwere implemented CL
=
cycle length (the time between two scheduled steam generator inspections)
VSL
=
structural limit voltage Gr
=
average growth rate per cycle length NDE
=
95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC)2 Implementation of these mid-cycle repair limits should follow same approach as in TS 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.
The NDE is the value provided by the NRC in GL 95-05.
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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 11.
F* Distance is the distance from the bottom of the hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.11 inches (not including eddy current uncertainty).
12.
Fd'ube is a tube'with degradation, below the F* distance, equal to or greater than 40%,
and not degiaded (i.e., no indications of cracking) within the F* distance.
13.
~Tube Re air refers to steevtng as described by the repons listed in 4.4.5.4.c which are used to maintain a tube in service or return a tube to service.
Tubes with degradation indications of less than the plugging limit may be preventively sleeved at the Owner's discretion.
This includes removal of plugs that were installed as a corrective or preventive measure.
A tube inspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service.
Further restrictions regarding identified indications and their proximity to the joint areas of various sleeving processes may be applicable.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plugging or sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.
Steam generator tube repairs may be made in accordance with the methods described in either WCAP-12623, WCAP-13088 (Rev. 3), or CEN-313-P.
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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 4.4.5.5 R~eo<<<
FoHowing each inservice inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the, number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.
The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.
This report shall include:
1.
Number and extent of tubes and sleeves inspected.
2.
Location and percent ofwall-thickness penetration for each indication ofan imperfection.
3.
Identification of tubes plugged or sleeved.
Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation.
The written followup of this 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 voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service should any of the following conditions arise:
If estimated leakage based on the projected end-of-cycle (or if-not practical, using the actual measured endwf-cycle) voltage distribution exceeds the leak limit(determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.
2.
If circumferential crack-like indications are detected at the tube support plate intersections.
Ifindications are identified that extend beyond the confines of the tube support plate.
If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
Ifthe calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured endwf cycle) voltage distribution exceeds I x 10, notify the Commission and provide an assessment of the safety significance of the occurrence.
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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM OPERATIONALLEAKAGE LIMITINGCONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARYLEAKAGE, b.
1 GPM UNIDENTIFIEDLEAKAGE, c.
600 gallons per day total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator, d.
10 GPM IDENTIFIED LEAKAGEfrom the Reactor Coolant System, Seal line resistance greater than or equal to 2.27E-1 ft/gpm~ and, The leakage from each Reactor Coolant System Pressure Isolation Valves specified in Table 3.44 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System average pressure within 20 psi of the nominal full pressure value.
APPLICABILITY MODES 1, 2, 3 and 4.*
ACTION:
With any PRESSURE BOUNDARYLEAKAGE, be in at least HOT STANDBYwithin 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 leakage greater than any one of the above limits, excluding PRESSURE BOUNDARYLEAKAGE,reduce the leakage rate to within limits within4 hours 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 />.'ith any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, declare the leaking valve inoperable and isolate the high pressure portion of the affected system from the low pressure portion by the use of a combination of at least two closed valves, one of which may be the OPERABLE check valve and the other a closed de-energized motor operated valve. Verifythe isolated condition of the closed de-energized motor operated valve at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-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 />.
Specification 3.4.6.2.e is applicable with average pressure within20 psi ofthe nominal fullpressure value.
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3/4 BASES 3/4,4 REACTOR COOLANT SYSTEM 1
3/4.4.5 STEAM GENERATORS TUBE INTEGRITY The Surveillance Requirements for inspection ofthe steam generator tubes ensure that the structural integrity ofthis portion of the RCS willbe 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 steain 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 willbe maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. Ifthe secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation ofsteam 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.
Axial or circumferentially oriented cracks having a primaty-to-secondary leakage less than this limitduring operation willhave an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Leakage in excess of this limit will require plant shutdown and an inspection, during which the leaking tubes willbe located and plugged or repaired.
A steam generator while undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/
operating upon reinstatement of auxiliary or main feed fiow control and steam control.
Wastage-type defects are unlikely with the all volatile treatment (AVT)of secondary coolant.
However, even if a defect of similar type should develop in service, it willbe found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the repair limit which is defined in Specification 4.4.5.4.a.
Steam generator tube inspections ofoperating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY Continued The voltage-based repair limits of these surveillance requirements (SR) implement the guidance in GL 9545 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections.
The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG. Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no NDE detectable cracks extending outside the thickness of the support plate.
Refer to GL 9545 for additional description of the degradation morphology.
Implementation of these SRs requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limitfrom the structural limit(which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure/bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650'F (i.e., the 95-percent LTLcurve).
The voltage structural limitmust be adjusted downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit;VURL, is determined from the structural voltage limitby applying the followingequation:
VURL = VSL - VGr - VNDE where VG, represents the allowance for degradation growth between inspections and VNDErepresents the allowance for potential sources oferror in the measurement ofthe bobbin coil voltage.
Further discussion of the assumptions necessaty to determine the voltage repair limit are discussed in GL 9545.
The mid-cycle equation in SR 4.4.5.4.10.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommended by GL 9545 for situations which the NRC wants to be notified prior to returning the SGs to service.
For the purposes of this reporting requirement, leakage
. and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle (EOC) voltage distribution (refer to GL 9545 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service.
Note that ifleakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.l and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should provided per the GL section 6.b (c) criteria.
Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criteria of 4.4.5.4.a.10.
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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY Continued Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption ofplant operation. Such cases willbe 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.
Degraded stcam generator tubes may be repaired by the installation of sleeves which span the section of degraded steam generator tubing. A steam generator tube with a sleeve installed meets the structural requirements of tubes'hich are not degraded.
To determine the basis for the sleeve plugging limit, the minimum sleeve wall thickness was calculated in accordance with Draft Regulatory Guide 1. 121 (August 1976) ~
In addition, a combined allowance of 20 percent of wall thickness is assumed for eddy current testing inaccuracies and continued operational degradation per Draft Regulatory Guide 1.121 (August 1976).
The following sleeve designs have been found acceptable by the NRC staff:
1.
Westinghouse Mechanical Sleeves (WCAP-12623) 2.
Combustion Engineering Leak Tight Sleeves (CEN-313-P) 3.
Westinghouse Laser Welded Sleeves (WCAP-13088, Rev. 3)
Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval in accordance with 10 CFR 50.90 prior to their use in the repair of degraded steam generator tubes.
The submittals related to other sleeve designs shall be made at least 90 days prior to use.
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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGEDETECTION 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.
3/4.4.6.2 OPERATIONAL 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 IDENTIFIEDLEAKAGElimitations provides allowance for a limited amount ofleakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The limitation on seal line resistance ensures that the seal line resistance is greater than or equal to the resistance assumed in the minimum safeguards LOCA analysis.
This analysis assumes that all ofthe flowthat is diverted from the boron injection line to the seal injection line is unavailable for core cooling.
Maintaining an operating leakage limitof 150 gpd per steam generator (600 gpd total) willminimize the potential for a large leakage event during steam line break under LOCA conditions.
Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate followinga steam line rupture is limited to below 8.4 gpm which willensure the calculated offsite doses willremain within 10 percent ofthe 10 CFR 100 requirements and that control room habitability continues to meet GDC-19. Leakage in the intact loops is limited to 150 gpd. If the projected end of cycle distribution of crack indications results in primary-to-secondary leakage greater than 8.4 gpm in the faulted loop during a postulated. steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 8.4 gpm.
Also, the 150 gpd leakage limit incorporated into this specification is more restrictive than the standard operating leakage limitand is intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or unexpectedly extend ouside the thickness of the the tube support plate.
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 willbe detected and the plant shut down in a timely manner.
PRESSURE BOUNDARYLEAKAGEofany magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Should PRESSURE BOUNDARYLEAKAGEoccur through a component
'which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.
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