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{{#Wiki_filter:ATTACHMENT 3 TO AEP:NRC:1166H MARKED-UP COPIES OF THE EXISTING T/S PAGES 9312270240 931215 PDR ADOCK 05000315 P             PDB
{{#Wiki_filter:ATTACHMENT 3 TO AEP:NRC:1166H MARKED-UP COPIES OF THE EXISTING T/S PAGES 9312270240 931215 PDR ADOCK 05000315 P
PDB


CTO   COO       SYSTEM SURVEILLANCE                   S   Continued
CTO COO SYSTEM SURVEILLANCE S
: 2. Tubes   in those   areas ~here experience has indicated   potential problems.
Continued Co 2.
: 3. h. tube inspection (pursuant to Specification 4.4.5.4.a.8)       shall be performed on each selected tube..       If any selected tube does not permit the passage of the eddy current probe for a tube
Tubes in those areas
                ~ inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
~here experience has indicated potential problems.
Co    The tubes selected as the second and third samples         (if required by Table 4.4>>2) during each insezvice inspection may be subjecte'd to a partial tube inspection provided:
3.
: l. The tubes     selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
: h. tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.. If any selected tube does not permit the passage of the eddy current probe for a tube
: 2. The inspections include thos'e     por'tions of the tubes where imperfections were previously       found..
~
l9"
inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
: d. Implementation of the steam generator tube/tube supp             plate interim plugging criteria for one fuel cycle (Cycle             requires a 100% 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 outer diameter stress'corrosion cracking     {ODSCC)   indications.
The tubes selected as the second and third samples (if required by Table 4.4>>2) during each insezvice inspection may be subjecte'd to a partial tube inspection provided:
The results of each sample inspection shall         be classified into one of the following three categories:
l.
CategoO           Ius ection Results C-l               Less than 5t of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
C-2               One   or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10%
d.
2.
The inspections include thos'e por'tions of the tubes where imperfections were previously found..
l9" Implementation of the steam generator tube/tube supp plate interim plugging criteria for one fuel cycle (Cycle requires a
100% 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 outer diameter stress'corrosion cracking
{ODSCC) indications.
The results of each sample inspection shall be classified into one of the following three categories:
CategoO Ius ection Results C-l Less than 5t 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.
of the total tubes inspected are degraded tubes.
C-3               Kore than 10% of the total. tubes inspected are degraded tubes or more than 1't of the inspected tubes are defective.
C-3 Kore than 10% of the total. tubes inspected are degraded tubes or more than 1't of the inspected tubes are defective.
COOK NUCLEAR PLANT       - UNIT 1           3/4 4-8               AMENDMENT NO. M, 454 166
COOK NUCLEAR PLANT - UNIT 1 3/4 4-8 AMENDMENT NO. M, 454 166


REACTOR COO     S STEMS SURVEILLANC     U         S   Co t ued around the U-bend to the top support of the cold leg. Por a tube in which the tube support plate elevation interim
REACTOR COO S
              , plugging limit has been applied,     the inspection will include all the hot leg intersections and       all'old   leg intersections down to, at least, the level of the last crack indication.
STEMS SURVEILLANC U
: 9.   ~aleevin a tahe is pernitaed only in areas vhere the sleeve spans the tubesheet area and whose lower joint is at the primary   fluid tubesheet   face.
S Co t ued around the U-bend to the top support of the cold leg.
: 10.       e Tube u ort late nterim Plu in Criteria is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates. Por application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude. The plant-specific guidelines used for all inspections shall'e amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters.     Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utili.red in the Donald C. Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.
Por a tube in which the tube support plate elevation interim
: 1. A tube can remain in service       if the signal amplitude of a crack indication is less than       or equal to 1.0 volt, regardless of the depth of tube wall penetration, if, as a result, the projected end-of-cycle distribution of
, plugging limit has been applied, the inspection will include all the hot leg intersections and all'old leg intersections down to, at least, the level of the last crack indication.
                    "'rack. indications is verified to result in'rimary-to-secondary leakage less than 1 gpm in the faulted loop during a postulated steam line break event. The methodology for calculating expected leak rates from the projected crack distribution must be consistent with VCAP-13187, Rev. 0)(~85 PgSi"AheQ 1ll Jfg jj jPg/Q /gP/
9.
: 2. h tube should be plugged or repaired if, the signal amplitude of the crack indication is greater than 1.0 volt except as noted in 4.4.5.4.a.10.3 below.
~aleevin a tahe is pernitaed only in areas vhere the sleeve spans the tubesheet area and whose lower joint is at the primary fluid tubesheet face.
: 3. A tube. can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than or equal Zdl     does volts if a rotating pancake probe inspection not detect degradation. Indications of degiadation with a bobbin coil signal amplitude greater than . volts will be plugged or repaired.
10.
                    ~,l COOK NUCLEAR PLANT   - UNIT   1       3/4 4-11              AMENDMENT NO. 08  454 166
e Tube u
ort late nterim Plu in Criteria is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates.
Por application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude.
The plant-specific guidelines used for all inspections shall'e amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters.
Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utili.red in the Donald C.
Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.
1.
A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volt, regardless of the depth of tube wall penetration, if, as a result, the projected end-of-cycle distribution of
"'rack. indications is verified to result in'rimary-to-secondary leakage less than 1 gpm in the faulted loop during a postulated steam line break event.
The methodology for calculating expected leak rates from the projected crack distribution must be consistent with VCAP-13187, Rev. 0)(~85 PgSi"AheQ 1llJfgjjjPg/Q /gP/
2.
h tube should be plugged or repaired if, the signal amplitude of the crack indication is greater than 1.0 volt except as noted in 4.4.5.4.a.10.3 below.
AMENDMENT NO. 08 454 166 3/4 4-11 3.
A tube. can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than or equal Zdl volts if a rotating pancake probe inspection does not detect degradation.
" Indications of degiadation with a bobbin coil signal amplitude greater than volts will be plugged or repaired.
~,l COOK NUCLEAR PLANT - UNIT 1


0   C G C ND         R p 3.4.6.2   Reactor Coolant System leakage shall be limited to:
0 C
: a. No PRESSURE BOUNDARY LEAKAGE,
G C ND R
: b. 1 GPH UNIDENTIFIED LEAKAGE, c     600 gallons pez day total pz         -to-secondazy leakage through all steam generators and     150'ns       per day through any one steam generaeor for cruel Cycle
p 3.4.6.2 Reactor Coolant System leakage shall be limited to:
: d. 10 GPH IDENTIFIED LE/QQLGE from     the Reactor Coolane System,
a.
: e. Seal line resistance greater than or     equal to 2.27E-1 ft/gpmi and, 1 GPH leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.
No PRESSURE BOUNDARY LEAKAGE, b.
AOTZON'ith any PRESSURE   BOUNDARY LEDGE;     be in ae lease HOT STANDBY within 6 hours and   in COLD SHUTDOWN   within the following 30 hours.
1 GPH UNIDENTIFIED LEAKAGE, c
: 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 hours or be in ae least HOT STANDBY within the next 6 hours and in COLD SHUTDOQH within the following 30 hours.
600 gallons pez day total pz
Ce     Kith any reactor coolane     system pressure isolation valve(s) leakage greater than the above     limit, except when:
-to-secondazy leakage through all steam generators and 150'ns per day through any one steam generaeor for cruel Cycle d.
I
10 GPH IDENTIFIED LE/QQLGE from the Reactor Coolane
: 1. The leakage   is less than or equal to 5.0 gpm, and
: System, e.
: 2. The most recene measured leakage does noe exceed the previous measured leakage* by an amount that reduces the
Seal line resistance greater than or equal to 2.27E-1 ft/gpmi and, 1 GPH leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.
* To   satisfy hD8h raquiremenes, measured leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved'rocedures and "supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
AOTZON'ith any PRESSURE BOUNDARY LEDGE; be in ae lease HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
Specifi.cation 3.4.6.2.e. is- applicable with average pressure within 20 .
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 hours or be in ae least HOT STANDBY within the next 6 hours and in COLD SHUTDOQH within the following 30 hours.
Ce Kith any reactor coolane system pressure isolation valve(s) leakage greater than the above limit, except when:
I 1.
The leakage is less than or equal to 5.0 gpm, and 2.
The most recene measured leakage does noe exceed the previous measured leakage* by an amount that reduces the AHENDHENT NO. kent 16jo Order dated April 20, 1981 To satisfy hD8h raquiremenes, measured leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved'rocedures and "supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
Specifi.cation 3.4.6.2.e. is-applicable with average pressure within 20.
psi of the nominal full pressure value.
psi of the nominal full pressure value.
COOK NUCLEAR PLANT     - UNIT 1       3/4 4-16                   AHENDHENT NO. kent 16jo Order dated April 20, 1981
COOK NUCLEAR PLANT - UNIT 1 3/4 4-16


0   C 4.4, The   Surveillance Requirements for inspection of the steam generator tubas ensure thae tha structural       integrity of this portion of ehe RCS vill be maintained.     The program for insaxvice inspection of steam genezator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
0 C
4.4, The Surveillance Requirements for inspection of the steam generator tubas ensure thae tha structural integrity of this portion of ehe RCS vill be maintained.
The program for insaxvice inspection of steam genezator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Xnservice inspe'ction of steam generator tubing is essential QL ordex to maintain surveillance of the conditions of the tubes in tha event that there is evidence of mechanical disaga or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Xnservice inspe'ction of steam generator tubing is essential QL ordex to maintain surveillance of the conditions of the tubes in tha event that there is evidence of mechanical disaga or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservica inspection of steam generator tubing also provides a, means 'of characterizing the natura and cause of any tube degradation so that corrective   measux'es   can be taken.
Inservica inspection of steam generator tubing also provides a, means 'of characterizing the natura and cause of any tube degradation so that corrective measux'es can be taken.
tq The plane   is   expected to be operated in a manner such tha the second-ary coolant   vill be   maintained vithin those chemistry limits ound to result in negligible   corrosion     of the steam generator tubes. If   e secondary coolane chemistry is not maintained vithin these paramete limits, localized cozrosion may likely result in stress corrosion cracking The extent of cracking during plant operation vould be limited by the imitation of steam generator tube leakage betveen the primary coolane syst and the secondary coolant system. The allovable primary-to-secondary le rate is 150 gallons per day per seeam generator for one fuel cycle (Cycle               . Axial or cizcumferentially oriented cracks having         a primary-to-secondary   leakage less than this limit during operation vill have an adequate margin of safety to vithstand ehe loads imposed during normal operation and by postulated accidents. Leakage in excess of this limit           vill require plane shutdovn and an inspection, during vhich the leaking tubes         vill  be located and plugged or repaired. A steam generator vhile undergoing crevice flushing in Node 4 is available for decay heat removal and is operable/opexaeing upon reinstatement of aux'.iazy or main feed flov control and steam control.
tq The plane is expected to be operated in a manner such tha the second-ary coolant vill be maintained vithin those chemistry limits ound to result in negligible corrosion of the steam generator tubes.
wastage-type defects ara unlikely       vith the all   volatQ.e treatment (AVT) of secondary coolant.       Hovevex', even if a defect of similar type should develop in service,       it vill   be found during scheduled insenrica steam vill be required for  all generator tube exanixmtions. Plugging or sleevtng tubes   vith imperfections exceeding the repair limit vhich,is defined in Specification 4.4.5.4.a. Steam generator tube inspections of operating plants have demonsezatad the capability to reliably detect degzadaeion that has penetrated 20% of the original tube vali thickness.
If e secondary coolane chemistry is not maintained vithin these paramete cozrosion may likely result in stress corrosion cracking cracking during plant operation vould be limited by the generator tube leakage betveen the primary coolane syst coolant system.
The allovable primary-to-secondary le limits, localized The extent of imitation of steam and the secondary rate is 150 gallons per day per seeam generator for one fuel cycle (Cycle Axial or cizcumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation villhave an adequate margin of safety to vithstand ehe loads imposed during normal operation and by postulated accidents.
Leakage in excess of this limit villrequire plane shutdovn and an inspection, during vhich the leaking tubes villbe located and plugged or repaired.
A steam generator vhile undergoing crevice flushing in Node 4 is available for decay heat removal and is operable/opexaeing upon reinstatement of aux'.iazy or main feed flov control and steam control.
wastage-type defects ara unlikely vith the all volatQ.e treatment (AVT) of secondary coolant.
Hovevex',
even if a defect of similar type should develop in service, it vill be found during scheduled insenrica steam generator tube exanixmtions.
Plugging or sleevtng villbe required for all tubes vith imperfections exceeding the repair limit vhich,is defined in Specification 4.4.5.4.a.
Steam generator tube inspections of operating plants have demonsezatad the capability to reliably detect degzadaeion that has penetrated 20% of the original tube vali thickness.
Tubes experiencing outer diameter stress corrosion czacking vithin ehe thickness of the tube support plates ara plugged or repaired by the criteria of 4.4.5.4.a.lO.
Tubes experiencing outer diameter stress corrosion czacking vithin ehe thickness of the tube support plates ara plugged or repaired by the criteria of 4.4.5.4.a.lO.
COOK NUCLEAR PLANT       - UNIT   1     B 3/4 4<<2a     hHEH".":~   NO. ~   444 166
COOK NUCLEAR PLANT -
UNIT 1 B 3/4 4<<2a hHEH".":~ NO. ~ 444 166


C 0   C   0 HaSntaining an operating     1     ge limit of 150 gpd per steam generator (600 gpd   total) for   Puel Cycle     vill minimi=e the potential fox' large leakage event duzing steam line reaR under LOCA conditions. Based on the NDE uncertainties, bobbin coil voltage distribution and crack grovth rate from the previous inspection, the expected leak rate folloving a steam line rupture is limited to belov 120 gpm in the faulted loop and 150 gpd per steam generator In the intact loops, vhich         vill limit offsite doses to vithin 10 percent of the 10 CPR 100 guidelines.           If the pro)ected end of cycle distzibution of czack indications results In         primary-to-secondary leakage greater than 120 gpm in the faulted loop during a postulated steam line break event, additional tubes must ba removed from service in order to reduce the postulated primary-to-secondazy steam line break leakage to belov 120 gpm.
C 0 C 0 HaSntaining an operating 1
PRESSURE   BOUNDARY LEAKAGE     of any magnitude is unacceptable since   it may be   indicative of   an Impending gross failure of the pressure boundary.
ge limit of 150 gpd per steam generator (600 gpd total) for Puel Cycle villminimi=e the potential fox' large leakage event duzing steam line reaR under LOCA conditions.
Based on the NDE uncertainties, bobbin coil voltage distribution and crack grovth rate from the previous inspection, the expected leak rate folloving a steam line rupture is limited to belov 120 gpm in the faulted loop and 150 gpd per steam generator In the intact loops, vhich villlimit offsite doses to vithin 10 percent of the 10 CPR 100 guidelines.
If the pro)ected end of cycle distzibution of czack indications results In primary-to-secondary leakage greater than 120 gpm in the faulted loop during a postulated steam line break event, additional tubes must ba removed from service in order to reduce the postulated primary-to-secondazy steam line break leakage to belov 120 gpm.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an Impending gross failure of the pressure boundary.
Should PRESSURE BOUNDARY UMQLGE occuz through a component vhich 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 souxca of potential failure.
Should PRESSURE BOUNDARY UMQLGE occuz through a component vhich 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 souxca of potential failure.
The Surveillance RequIrements for RCS Pressure Xsolation 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 ZDENTXPXED LEAKAGE and         vill be considered as a portion of the alloved 11mit.
The Surveillance RequIrements for RCS Pressure Xsolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
3   4.4.7 CKBfISTRY The   limitations   on Reactor Coolant   Syst~ chemistry ensure that corrosion of the Reactor Coolant         System is minimized and reduces the potential for     Reactor Coolant System leakage or failure due to stress corrosion. maintaining the chemistry vithin the Steady State Limits provides adequate corrosion protection to ensuze the structural integrity of the Reactor Coolant System ovei the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are'time and temperature dependent. Corrosion studies shov that operation may'e continued vith contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, fox the specified limited time intervals vithout having a signIficant effect on the structural integrity of the Reactor Coolant System. The time Interval permitting continued operation vithin the restrictions of the Transient Limits provides time for taking corrective actIons to restore the contaminant concentrations to vithin the Steady State Limits.
Leakage from the RCS Pressure Isolation Valves is ZDENTXPXED LEAKAGE and villbe considered as a
COOK NUCLEAR PLANT     - UHXT 1       B 3/4 4-4                 hHEHDHZNT NO   M 166.
portion of the alloved 11mit.
3 4.4.7 CKBfISTRY The limitations on Reactor Coolant Syst~ chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
maintaining the chemistry vithin the Steady State Limits provides adequate corrosion protection to ensuze the structural integrity of the Reactor Coolant System ovei the life of the plant.
The associated effects of exceeding the oxygen, chloride, and fluoride limits are'time and temperature dependent.
Corrosion studies shov that operation may'e continued vith contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, fox the specified limited time intervals vithout having a signIficant effect on the structural integrity of the Reactor Coolant System.
The time Interval permitting continued operation vithin the restrictions of the Transient Limits provides time for taking corrective actIons to restore the contaminant concentrations to vithin the Steady State Limits.
COOK NUCLEAR PLANT - UHXT 1 B 3/4 4-4 hHEHDHZNT NO M 166.


ATTACHMENT 4 TO AEP:NRC:1166H PROPOSED REVISED T/S PAGES}}
ATTACHMENT 4 TO AEP:NRC:1166H PROPOSED REVISED T/S PAGES}}

Latest revision as of 14:18, 7 January 2025

Proposed Tech Specs to Allow Unit 1 SG Tube Interim Plugging Criteria of 1.0 Volt for Cycle 14
ML17331B139
Person / Time
Site: Cook American Electric Power icon.png
Issue date: 12/15/1993
From:
INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17331B140 List:
References
NUDOCS 9312270240
Download: ML17331B139 (7)
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Text

ATTACHMENT 3 TO AEP:NRC:1166H MARKED-UP COPIES OF THE EXISTING T/S PAGES 9312270240 931215 PDR ADOCK 05000315 P

PDB

CTO COO SYSTEM SURVEILLANCE S

Continued Co 2.

Tubes in those areas

~here experience has indicated potential problems.

3.

h. tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.. If any selected tube does not permit the passage of the eddy current probe for a tube

~

inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

The tubes selected as the second and third samples (if required by Table 4.4>>2) during each insezvice inspection may be subjecte'd to a partial tube inspection provided:

l.

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

d.

2.

The inspections include thos'e por'tions of the tubes where imperfections were previously found..

l9" Implementation of the steam generator tube/tube supp plate interim plugging criteria for one fuel cycle (Cycle requires a

100% 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 outer diameter stress'corrosion cracking

{ODSCC) indications.

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

CategoO Ius ection Results C-l Less than 5t 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 Kore than 10% of the total. tubes inspected are degraded tubes or more than 1't of the inspected tubes are defective.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-8 AMENDMENT NO. M, 454 166

REACTOR COO S

STEMS SURVEILLANC U

S Co t ued around the U-bend to the top support of the cold leg.

Por a tube in which the tube support plate elevation interim

, plugging limit has been applied, the inspection will include all the hot leg intersections and all'old leg intersections down to, at least, the level of the last crack indication.

9.

~aleevin a tahe is pernitaed only in areas vhere the sleeve spans the tubesheet area and whose lower joint is at the primary fluid tubesheet face.

10.

e Tube u

ort late nterim Plu in Criteria is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates.

Por application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude.

The plant-specific guidelines used for all inspections shall'e amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters.

Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utili.red in the Donald C.

Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.

1.

A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volt, regardless of the depth of tube wall penetration, if, as a result, the projected end-of-cycle distribution of

"'rack. indications is verified to result in'rimary-to-secondary leakage less than 1 gpm in the faulted loop during a postulated steam line break event.

The methodology for calculating expected leak rates from the projected crack distribution must be consistent with VCAP-13187, Rev. 0)(~85 PgSi"AheQ 1llJfgjjjPg/Q /gP/

2.

h tube should be plugged or repaired if, the signal amplitude of the crack indication is greater than 1.0 volt except as noted in 4.4.5.4.a.10.3 below.

AMENDMENT NO. 08 454 166 3/4 4-11 3.

A tube. can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than or equal Zdl volts if a rotating pancake probe inspection does not detect degradation.

" Indications of degiadation with a bobbin coil signal amplitude greater than volts will be plugged or repaired.

~,l COOK NUCLEAR PLANT - UNIT 1

0 C

G C ND R

p 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPH UNIDENTIFIED LEAKAGE, c

600 gallons pez day total pz

-to-secondazy leakage through all steam generators and 150'ns per day through any one steam generaeor for cruel Cycle d.

10 GPH IDENTIFIED LE/QQLGE from the Reactor Coolane

System, e.

Seal line resistance greater than or equal to 2.27E-1 ft/gpmi and, 1 GPH leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.

AOTZON'ith any PRESSURE BOUNDARY LEDGE; be in ae lease HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY 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 ae 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 SHUTDOQH within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Ce Kith any reactor coolane system pressure isolation valve(s) leakage greater than the above limit, except when:

I 1.

The leakage is less than or equal to 5.0 gpm, and 2.

The most recene measured leakage does noe exceed the previous measured leakage* by an amount that reduces the AHENDHENT NO. kent 16jo Order dated April 20, 1981 To satisfy hD8h raquiremenes, measured leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved'rocedures and "supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Specifi.cation 3.4.6.2.e. is-applicable with average pressure within 20.

psi of the nominal full pressure value.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-16

0 C

4.4, The Surveillance Requirements for inspection of the steam generator tubas ensure thae tha structural integrity of this portion of ehe RCS vill be maintained.

The program for insaxvice inspection of steam genezator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Xnservice inspe'ction of steam generator tubing is essential QL ordex to maintain surveillance of the conditions of the tubes in tha event that there is evidence of mechanical disaga or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservica inspection of steam generator tubing also provides a, means 'of characterizing the natura and cause of any tube degradation so that corrective measux'es can be taken.

tq The plane is expected to be operated in a manner such tha the second-ary coolant vill be maintained vithin those chemistry limits ound to result in negligible corrosion of the steam generator tubes.

If e secondary coolane chemistry is not maintained vithin these paramete cozrosion may likely result in stress corrosion cracking cracking during plant operation vould be limited by the generator tube leakage betveen the primary coolane syst coolant system.

The allovable primary-to-secondary le limits, localized The extent of imitation of steam and the secondary rate is 150 gallons per day per seeam generator for one fuel cycle (Cycle Axial or cizcumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation villhave an adequate margin of safety to vithstand ehe loads imposed during normal operation and by postulated accidents.

Leakage in excess of this limit villrequire plane shutdovn and an inspection, during vhich the leaking tubes villbe located and plugged or repaired.

A steam generator vhile undergoing crevice flushing in Node 4 is available for decay heat removal and is operable/opexaeing upon reinstatement of aux'.iazy or main feed flov control and steam control.

wastage-type defects ara unlikely vith the all volatQ.e treatment (AVT) of secondary coolant.

Hovevex',

even if a defect of similar type should develop in service, it vill be found during scheduled insenrica steam generator tube exanixmtions.

Plugging or sleevtng villbe required for all tubes vith imperfections exceeding the repair limit vhich,is defined in Specification 4.4.5.4.a.

Steam generator tube inspections of operating plants have demonsezatad the capability to reliably detect degzadaeion that has penetrated 20% of the original tube vali thickness.

Tubes experiencing outer diameter stress corrosion czacking vithin ehe thickness of the tube support plates ara plugged or repaired by the criteria of 4.4.5.4.a.lO.

COOK NUCLEAR PLANT -

UNIT 1 B 3/4 4<<2a hHEH".":~ NO. ~ 444 166

C 0 C 0 HaSntaining an operating 1

ge limit of 150 gpd per steam generator (600 gpd total) for Puel Cycle villminimi=e the potential fox' large leakage event duzing steam line reaR under LOCA conditions.

Based on the NDE uncertainties, bobbin coil voltage distribution and crack grovth rate from the previous inspection, the expected leak rate folloving a steam line rupture is limited to belov 120 gpm in the faulted loop and 150 gpd per steam generator In the intact loops, vhich villlimit offsite doses to vithin 10 percent of the 10 CPR 100 guidelines.

If the pro)ected end of cycle distzibution of czack indications results In primary-to-secondary leakage greater than 120 gpm in the faulted loop during a postulated steam line break event, additional tubes must ba removed from service in order to reduce the postulated primary-to-secondazy steam line break leakage to belov 120 gpm.

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

Should PRESSURE BOUNDARY UMQLGE occuz through a component vhich 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 souxca of potential failure.

The Surveillance RequIrements for RCS Pressure Xsolation 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 ZDENTXPXED LEAKAGE and villbe considered as a

portion of the alloved 11mit.

3 4.4.7 CKBfISTRY The limitations on Reactor Coolant Syst~ chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.

maintaining the chemistry vithin the Steady State Limits provides adequate corrosion protection to ensuze the structural integrity of the Reactor Coolant System ovei the life of the plant.

The associated effects of exceeding the oxygen, chloride, and fluoride limits are'time and temperature dependent.

Corrosion studies shov that operation may'e continued vith contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, fox the specified limited time intervals vithout having a signIficant effect on the structural integrity of the Reactor Coolant System.

The time Interval permitting continued operation vithin the restrictions of the Transient Limits provides time for taking corrective actIons to restore the contaminant concentrations to vithin the Steady State Limits.

COOK NUCLEAR PLANT - UHXT 1 B 3/4 4-4 hHEHDHZNT NO M 166.

ATTACHMENT 4 TO AEP:NRC:1166H PROPOSED REVISED T/S PAGES

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