ML20135F310
| ML20135F310 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/26/1996 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML19310D744 | List: |
| References | |
| NUDOCS 9612120492 | |
| Download: ML20135F310 (18) | |
Text
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PROPOSED TECHNICAL SPECIFICATION CHANGES i
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9612120492 961126 PDR ADOCK 05000368 P
PDR 1
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3 SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance criteria j
a.
As used~in this Specification t
1.
Tubing or Tube means that portion of the tube or sleeve which
[
forms the primary system to secondary system pressure boundary.
5 2.
Imperfection means an exception to the dimensions, finish or l
contour of a tube from that required by fabrication drawings j
or specifications.
Eddy-current testing indications below
(
20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
t j
3.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside I
of a tube.
i j
4.
Degraded Tube means a tube containing imperfections 220% of nominal wall thickness caused by degradation.
1 5.
% Degradation means the percentage of the tube wall j;
thickness affected or removed by degradation.
1 6.
Defect means an imperfection of such severity that it j
exceeds the plugging or repair limit. A tube containing 1
a defect is defective.
t 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 l
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 29%
of the nominal sleeve wall thickness within the sleeve pressure boundary for sleeves installed in accordance with CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight Sleeves," Revision 01, dated November 1996.
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 i
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 j
tube from the point of entry (hot leg side) completely
~
around the U-bend to the top support of the cold leg.
1 ARKANSAS - UNIT 2 3/4 4-9 Amendment No. 443,443,
~
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 CPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.
Defective tubes may be repaired in accordance with.
l 1)
B&W Topical Report BAW-2045PA-00 as supplemented by the j
information provided in B&W Report 51-1212539-00, "BW:IS Kinetic l
Sleeve Design-Application to ANO Unit 2".
2)
CENO Report CEN-630-P, " Repair of 3/4" O.D.
Steam Generator Tubes Using Leak Tight Sleeves," Revision 01, dated November 1996. The post weld heat treatment described in CEN-630-P shall be performed.
)
4.4.5.5 Reports j
l 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 Conadssion 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.
1 3.
Identification of tubes plugged or sleeved.
c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Jpecial 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.
ARKANSAS - UNIT 2 3/4 4-10 Amendment No. 91,443,444,
REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited tot a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE,.
300 gallons per 'ay total primary-to-secondary leakage through both l
d c.
steam generators and 150 gallons per day through any one steam generator, i
l 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and j
e.
Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6.1.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY 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 />.
c.
With any Reactor Coolant System Pressure Isolation Valve leakage 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.
Order dated 0/20/S1
{
BASES Demonstration of the safety valves' lift setting will occur only i
during shutdown and will be performed in accordance with the j.
provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
J i
l 3/4.4.4 PRESSURIZER 2
1 A steam bubble.in the pressurizer ensures that the RCS is not a i
hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code 4
safety valves against water relief. The steam bubble functions to relieve RCS pressure during all design transients.
d I
l The requirement that 150 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an i
emergency bus provides assurance that these heaters can be energized during j
a loss-of-offsite power condition to maintain natural circulation at HOT STANDBY.
1 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator l
tubes ensure that the structural integrity of this portion of the RCS will
{
be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to j
maintain surveillance of the conditions of the tubes in the event that i
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 j
a means of characterizing the nature and cause of any tube degradation l
so that corrective measures can be taken.
j The plant is expected to be operated in a manner such that the
)
i secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the 4
j 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 per steam generator).
Cracks having a primary-to-l i
secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation 1
and by postulated accidents. Operating plants have demonstrated that 5
primary-to-secondary leakage of 150 gallons per day per steam generator i
can readily be detected by radiation monitors on the secondary system.
i Leakage in excess of this limit will require plant shutdown and an unscheduled i
inspection, during which the leaking tubes will be located and plugged or j
repaired.
4 i
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ARKANSAS - UNIT 2 B 3/4 4-2 Amendment No. 40,MS, J
L REACTOR COOLANT OYSTEM i
BASES l
Wastage-type defects are unlikely with proper chemistry treatment of 4
the secondary coolant. However, even if a defect should develop in service, i
i 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 j
to ANO Unit 2" or CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight' Sleeves," Revision 01, dated November 1996.
Steam 3
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 the system that is used for periodic inservice inspection will be validated. Additionally, upgraded testing methods will be evaluated and appropriately implemented as better i
methods are developed and validated for commercial use.
J 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 t
of the Commission will be made prior to resumption of plant operation. Such cases will be considered by the Conadssion on a case-by-case basis and may result in a j
requirement for analysis, laboratory examinations, tests, additional eddy-current J
inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6' REACTOR COOLANT SYSTEM LEAKAGE l
3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systens 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 i
of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage j
Detection Systems" May 1973, a
l 3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE 5
Industry experience has shown.that while a limited amount of leakage l
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 d
Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
1 3-i ARKANSAS - UNIT 2 B 3/4 4-3 Amendment No. 94,443,44G, 9:d:: d: :d 4 2C S1
t REACTOR COOLANT SYSTEM BASES F
The total steam generator tube leakage limit of 300 gallons per day l
l for all steam generators ensures that the dosage contribution from t
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 150 gallon per day leakage limit per steam generator ensures that steam generator tube integrity is naintained in the event of a main steam line rupture i
or under LOCA conditions.
s PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since
}
it may be indicative of an impending gross failure of the pressure i
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 minindzed and reduce the t
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 4
continued with contaminant concentration levels in excess of the steady j
State Limits, up to the Transient Limits, for the specified limited time j
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 f
for taking corrective actions to restore the contaminant concentrations j ~
to within the Steady State Limits.
1 i
The surveillance requirements provide adequate assurance that con-centrations in excess rf the limits will be detected in sufficient time to take corrective action.
}
3/4.4.8 SPECIFIC ACTIVITY j
The limitations on the specific activity of the primary coolant 3
ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a 4
4 S
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1 ij ARKANSAS - UNIT 2 B 3/4 4-4 Amendment No.
)
l MARKUP OF CURRENT ANO-2 TECHNICAL SPECIFICATIONS 4
(FOR INFORMATION ONLY)
I e
d
+
S REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
I l
4.4.5.4 Acceptance Criteria a.
As used in this Specification
)
1.
Tubing or Tubg 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.
I 3.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside j
of a tube.
4.
Degraded Tube means a tube containing imperfections 220% of j
nominal wall thickness caused by degradation.
I 5.
% Degradation means the percentage of the tube wall 1
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.
l 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 i
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 -
l Application to ANO Unit 2".
The plugging limit is equal to g),44% of the nominal sleeve wall thickness _wlthin the sierv3 oressure boundarv for sleeves installed in accordance with CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator Tubes Usino Leak Ticht Sleeves," Revision 01, dated November 1996a.CO!: ".;p :t 00: 501 ",
".'_"O-2 St;;; C;.. ::t : "th Z:pci
;ing L;;h Tight Sl;;;;;", ";;i;i;n 01-P, d;t;d July,1902.
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 gerarator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
ARKANSAS - UNIT 2 3/4 4-9 Amendment No. -143,443,
t 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 i
current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed I
after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) 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)
CENO Report CEN-630-P. "Renair of 3/4" O.D. Steam Generator Tubes Usina Leak Tiaht Sleeves." Revision 01. dated November 1996. The oost weld heat treatment described in CEN-630-P shall be nerformed.
CD?r n p :t 00: 501-T, "?J:0 0 Ot := 0:ne::::: Td nep;ir 'J:ing L :h Tight 01 ::::", n:vici:n 01-",
d:ted 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.
i 3.
Identification of tubes plugged or sleeved.
c.
Results of steam generator tube inspections which fall 4.nto Category C-3 shall be reported in a Special Report p ruuant to Specification 6.9.2 as denoted by Table 4.4-2.
Notin ra; ion of the Commission will be made prior to resumption of plant operation. The written Special Report shall provide a cescription of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
j ARKANSAS - UNIT 2 3/4 4-10 Amendment No. M,443,443,
l 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.
300 callons per dav4-GPM total primary-to-secondary leakage through both steam generators and 150 aallons per dav0.5 CPM through any one steam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.
Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6.1.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY 4
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 Sysiim leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within lindts 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 i
3 the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
I c.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion i
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.
j d
i 1
J ARKANSAS - UNIT 2 3/4 4-14 Amendment No.
Cri;; i:tzi ^!:Cl?1 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 i
STANDBY.
3/4.4.5 STEAM GENERATORS i
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage
= 150_ gallons per dav0.5 CPM per steam generator).
Cracks having a primary-to-l 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. Operating plants have demonstrated that primary-to-secondary leakage of 150 callons per dav0.5 CPM per steam generator can readily be detected by radiation monitors on the sec9ndary systemof steam g.;;rctee-blewdown.
Leakage in excess of this limit will require plant shutdown and an unsche: led inspection, during which the leaking tubes will be located and p).sged or repaired.
ARKANSAS - UNIT 2 B 3/4 4-2 Amendment No. 24,MS,
+
1 i
i REACTOR COOLANT SYSTEM BASES U
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 CENO Report CEN-630-P.
- Repair of 3/4" O.D.
Steam Generator Igbes Usino Leak Tiaht Sleeves," Pevision 01, dated November 1996 GEN nep :t CEN-5 01 --- 0, "."."O -- 2 S t e :- Ce nc eat e r T d n:pcir U:ing Lech Tight Sleeves", Revision Ob Br-dabed-July 1002.
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 the system that is used for periodic inservice inspection will be validated. Additionally, l
upgraded testing methods will be evaluated and appropriately implemented as better methods are developed and validated for commercial use.
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 Conadssion 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.
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 1
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.
ARKANSAS - UNIT 2 B 3/4 4-3 Amendment No. &1,444,444, C:d:: d:ted
- D 91
REACTOR COOLANT SYSTEM BASES The total steam generator tube leakage limit of 300 callons per day 4-GPM for all steam generators 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 CP" kimit i: cen;istent with th :::u=ption: used in th: :n:1 :i: Of th:::
j
- id:nt:. The 150 callon per dav0.5 CP." leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main.
steam line rupture or under LOCA conditions.
PRESSURE BOUNDAT.Y LEAKAGE of any magnitude is unacceptable since it may be indicative of an inpending 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 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 con-centrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY 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 lindts following a
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i ARKANSAS - UNIT 2 B 3/4 4-4 Amendment No.
AFFIDAVIT PURSUANT TO 10 CFR 2.790 I, Ian C. Rickard, depose and say that I am the Director, Operations Licensing, of Combustion Engineering, Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conjunction with the application of Entergy Operations, incorporated and in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations.
The information for which proprietary treatment is sought is contained in the following document:
CEN-630-P, Rev. 01, " Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight Sleeves," November 1996.
This document has been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial information.
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.
1.
The information sought to be withheld from public disclosure, is owned and i
has been held in confidence by Combustion Engineering. It consists of the details concerning the fabrication process, material properties, and
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' surveillance data used to develop an approach to ascertain the embrittlement of reactor vessels.
2.
The information consists of test data or other similar data concerning a process, method or component, the application of which results in substantial competitive advantage to Combustion Engineering.
3.
The information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter DP-537 from F. M. Stern to Frank Schroeder dated December 2,1974. This system was applied in determining that the subject document herein is proprietary.
4.
The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.
5.
The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
6.
Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because:
a.
A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion Engineering.
b.
Development of this information by Combustion Engineering required tens of thousands of dollars and thousands of j
' manhours of effort. A competitor would have to undergo similar expense in generating equivalent information.
c.
In order to acquire such information, a competitor would also require considerable time and inconvenience to develop the design, development, and installation process for a welded sleeve for repairing 3/4 inch O.D. steam generator tubes.
d.
The information consists of the design, development, and installation process for a welded sleeve for repairing 3/4 inch j
O.D. steam generator tubes, the application of which provides a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take 2
marketing or other actions to improve their product's position or impair the position of Combustion Engineering's product, and avoid developing similar data and analyses in support of their l
processes, methods or apparatus.
e.
In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and expenses must be included. The ability of Combustion Engineering's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.
f.
Use of the information by competitors in the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with their technology development In addition, disclosure would have an adverse economic impact on Combustion 4
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4_
Engineering's potential for obtaining or maintaining foreign licensees.
i Further the deponent sayeth not.
/0 lanb Rick Director, Operations Licensing 4
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Sworn to before me this / 9 N day of
>/A J
.1996 V
Notary Public
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My. commission expires: 231 99
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