ML20137R594
| ML20137R594 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/09/1997 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137R599 | List: |
| References | |
| NUDOCS 9704140181 | |
| Download: ML20137R594 (26) | |
Text
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t UNITED STATES
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NUCLEAR REGULATORY COMMISSION 2.
' WASHINGTON, D.C. NMi6&4001
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SE000YAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 222 License No..DPR-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated October 18, 1996 as supplemented March 12, March.17, April 4 and April 9,1997, complies with the standards and l
requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;
-C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and i
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 1
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
I l
9704140181 970409 W
PDR ADOCK 05000327 P
PDR p-
2.
Accordingly, the license is amended by changes to-the Technical Specifications as indicated in the attachment to this license amendment.
. Paragraphs 2.C.(2) and 2.C.(9) of Facility Operating License No. DPR-77 i
are hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 222, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
A~new condition 2.C.(9)(d) is added as follows:*
(9) Steam Generator Insoection (Section 5.3.11
( d,'
By May 20, 1997, TVA shall establish a steam ge..erator inspection program that is in accordance with the commitments listed in Enclosure 2 to the TVA letter to the Commission on this subject dated March 12, 1997.
3.
This license amendment is effective as of its date of issuance, to be implemented no later than 45 days of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION g.d s A e,$l~! j.]1L, Frederick J. Hebdon, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
1.
Page 5 of License 2.
Changes to the Technical Specifications Date of Issuance: April 9,1997
- Page 5 of the composite license is attached to reflect this change L
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(9)-
Steam Generator' Inspection (Section 5.3.1)
(a)
Prior to March 1, 1981, TVA shall provide to the NRC' the results of its tests to determine the feasibiity l
of using a steam generator camera device.
(b)
Prior to start-up after the first refueling, TVA must install inspection ports in each steam generator if the results of the camera device inspection are not l
satisfactory to the NRC.
l l
(c)
Prior to start-up after the first refueling,.TVA will plug Row 1 of the steam generator tubes, if required by NRC.
(d)
By May 20, 1997, TVA shall establish a steam generator inspection program that is in accordance with the commitments listed in Enclosure 2 to the TVA letter to the Commission on this subject dated March 12, 1997, as modified by TVA letter dated March 17, 1997.
(10)
Water Chemistry Control Proaram (Section 5.3.2)
This requirement has been deleted.
(11)
Neoative Pressure in the Auxiliary Buildino Secondary Containment Enclorre (ABSCE) (Section 6.2.3)
After the final ABSCE configuration is determined, TVA must demonstate to the satisfaction of the NRC that a negative pressure of 0.25 inches of water gauge can be maintained in
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'the spent fuel storage area and in the ESF pump room.
]
l (12)
Environmental Oualification (Section 7.2.2)
(a)
No later than November 1,1980, TVA shall submit
)
i-information to show compliance with the requirements of NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related j
Elecctrical Equipment," for safety-related equipment exposed to a harsh environment.
Implementation shall be in accordance with NUREG-0588 by June 30, 1982.
(b)
By no later than December 1, 1980, complete and j
auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DDR Guidelines or NUREG-0588.
Thereafter,-such records should be updated and maintained current as equipment is 1
' replaced, further tested, or otherwise further qualified to document complete compliance by
)
June 30, 1982.
Amendment No. 222 1
ATTACHMENT TO LICENSE AMENDMENT NO. 222 i'
FACILITY OPERATING LICENSE N0. DPR-77 DOCKET NO. 50-327 Revise the Appendix A' Technical Specifications by removing the pages
-i
- identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal-lines indicating the area of change, REMOVE INSERT 3/4 4-7
'3/4 4-7' i
3/4 4-9 3/4 4-9 3/4 4-9a 3/4.4-9a j
3/4 4-9b 3/4 4-9b 3/4 4-10 3/4 4-10 3/4 4-14 3/4 4-14 8 3/4 4-3 B 3/4 4-3 l
B 3/4 4-4 8 3/4 4-4 j
B 3/4 4-4a B 3/4 4-4a r
i t
h b
1 4
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f i
s i
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d
REACTOR C00LANT SYSIB SURVEILLANCE REQUIREMENTS (Continued) i
'3.
-A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall l
be performed on each selected tube.
If any selected tube does.
i -
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.
I 4.
Indications left in service as a result of ap lication of the tube support' plate voltage-based repair criteria skall be inspected by l
. bobbin coil probe during all future refueling outages.
c.
The tubes selected as the second and third samples (if required by j
Table 4.4-2) during each inservice inspection may be subjected to a 4
partial tube inspection provided:
1.
The tubes selected for Wse samples include the tubes from those areas of the tube sheet array whero tubes with' imperfections were previously found.
l 2.
The inspections include those portions of the tubes ishore
)
imperfections were previously found.
l NOTE:
Tube degradation identified in the portion of the tube that is not a reactor coolant pressure boundary (tube end up to gig:
the start of the tube-t.o-tubesheet weld) is excluded from the Result and Action Required fn Table 4.4-2.
l d.
Implementation of the steam generator tube / tube suppori; plate repair i
criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg i
tube support plate with knoun outside diameter stress correzion cracking (00 SCC) indications. The determination of the lowest cold-leg I
i tube support plate intersections having 005CC indications shall be l
based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.
The results of each sample inspection shall be classified into one of the following three categories:
1 Catenary Insnection Results i
C-1 Less than 55 of the total tubes inspected are degraded l
tubes and none of the inspected tebes are defect' vs.
C-2 One or more tubes, but not more than 15 of the total i
tubes inspected are defective, or between 55 and 105 of the total tubes inspected are degraded tubes.
C-3 More than 105 of the total tubes inspected are degraded tubes or more than 15 of the inspected tubes are defective.
l Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 105) further wall penetrations to be included in the above percentage calculations.
j.
SEQUOYAH - UNIT 1 3/4 4-7 Amendment No.18g, 214,222
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REACTOR CCOLRNT SYSIEM F
SURVEILLANCE REQUIREMENTS (Continued)
---e.
_. =r_s 4.4.5.4 Acceptance Critaria a.
As used in this Specification:
- 2..
Jmoerfection means an exception to the dimensions, finish or contaur 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 he considered as imperfections.
2.
LeoradaLign means a service-induced cracking, wastage, wear or general corrosion occuring or either inside or outside of a tube.
3.
Everaded Tube means a tube containirag imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.
t Deoradation means the percenta9e of the tube wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective.
6.
Pluccina Lim 41 means the imperfection depth at or beyond which the tube atall be removed from service and is equal to 40% of the nominal tube wall thickness.
Plugging limit does R193 not apply to that portion of the tube that is not within the pressure boundary of the reactor coolant system (tube end up to the start of the tube-to-tubesheet weld).
This definition does not apply to tube support plate intersections if the voltage-based repair criteria are being applied.
Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.
7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large wnough 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.
8.
Tube Inscection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the. cold leg.
9.
Preservice Inspection means a tube inspection of the full length of each tube in each steam generator perirrmed by eddy current techniques prior to service establish a baseline con-dition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10.
Tube Sucoort Plate Pluccino Limit is used for the disposition of an alloy 600 steam generator tube for continued service that le experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the SEQUOYAH - UNIT 1 3/4 4-9 Amendment No. 189, 214, 222
i J
(Continued).
SURVEILLANCE REQUIREMENTS 1
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:
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a.
Steam generator 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 the lower voltage repair.
limit (Note 1), will be allowed to remain in service.
b.
Steam generator tubes, whose degradation is attributed j
to outside diameter stress corrosion cracking within t.he bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1), will be repaired or plugged, except as noted in 4.4.5.4.a.10.c bulow.
c.
Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion-cracking within the bounds of the tube
.i support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1), but less than or equal to upper voltage repair limit (Note 2), may remain in service if a rotating pancake coil inspection i
does not detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion-cra cking degradation with a bobbin coil voltage greater i
than the upper voltage repair limit (Note 2) will be I
plugged or repaired, d.
Not applicable to SQN.
e.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.
The mid-cycle repair limits are determined from the following equations.
v3t
- " ' ~
(CL-at) 1.0+NDE+Gr CL J
(CL-At) l uunt- (v t-V at) v tat.v l
un t
u
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\\
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i SEQUOYAH - UNIT 1 3/4 4-9a Amendment No. 189, 214, 222 1
__r e
i ESCTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) where:
upper voltage repair limit V
=
un.
lower voltage repair limit V u.
=
L mid-cycle upper voltage repair limit based on Vuma time into cycle mid-cycle lower voltage repair limit based on Vuutt
=
Vwust and time into cycle length of time since last scheduled inspection At
=
and V at were implemented during which Vunt t
cycle length (the time between two scheduled CL
=
steam generator inspections) structural limit voltage V
=
3t average growth rate per cycle length q
Gr
=
95-percent cumulative probability allowance for NDE
=
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.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.
Note 1:
The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing or 2.0 volts for 7/8-inch diameter tubing.
Note 2:
The upper voltage repair limit is calculated according to the j
methodology in GL 90-05 as supplemented.
Veng may differ at the TSPs and flow distribution baffle.
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SEQUOYAH - UNIT 1 3/4 4-9b Amendment No, 189, 214,222
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.
The steam genezator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by l
Table 4.4-2.
4.4.5.5 Reoorts a.
Following each inservice inspection of steam generator tubes, the number of tubes plugged 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 submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following completion of the inspection. This Special Report shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged, c.
Results of steam generator tube inspections which fall into lR40 Category C-3 shall be reported pursuant to Specification 6.6.1 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.
d.
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returaing the steam generators to service should any of the following 1
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 steam line break) for the next operating cycle.
2.
If circumferential crack-like indications are detected at the tube support plate intersections.
I 3.
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.
5.
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 108, i
notify the NRC and provide an assessment of the safety significance of the occurrence.
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l SEQUOYAH - UNIT 1 3/4 4-10 Amendment No. 36, 214, 222
)
REACTOR COOLANT SYSTEM OPEPATIONAL 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.
150 gallons per day of primary-to-secondary leakage through any one steam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.
40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig.
f.
1 GPM leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in R16 Table 3.4-1.
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 SHITTDOWN 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, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage l
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 l
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 j
hours.
j c.
With any Reacter 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 closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS l
4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within R16 Gach of the above limits by:
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l SEQUOYAH - UNIT 1 3/4 4-14 Amendment No. 12, 214,222
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REACTOR COOLANT SYSTEM EASES
.n 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 (prima ry-to-secondary leakage = 150 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Sequoyah has demonstrated that primary-to-secondary leakage of 150 gallons per day steam generator can readily be detected by radiation monitors of steam generator blowdown or condenser off-gas.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
The voltage-based repair limits of SR 4.4.5 implement the guidance in GL-95-05 and are applicable only to Westinghouse-designed steam generators (S/Gs) 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 S/G tube degradation nor are they applicable to ODSCC that occurs at other locations within the S/G. Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the thickness of the support plate. Refer to GL 95-05 for additional description of the degradation morphology.
Implementation of SR 4.4.5 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'to 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 flaw growth during an operating interval and to account for NDE uncertainty. The upper v,t, is determined from the structural voltage limit by voltage repair limit; u
applying the following equation:
v
-vn - Vc, - Vuos ung where Vca represents the allowance for flaw growth between inspections and Ves 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.
SEQUOYAH - UNIT 1 B 3/4 4-3 Amendment No. 36, 189, 214, 222
s REACTOR COOLANT SYg. TIM n
BASES
. The'mid-cycle' equation of'SR 4.4.5.4'.s.10.e=should only be used during unplanned' inspection in which eddy current data is acquired for indications at r
l
- the tube support. places.-
l 1
SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 forcsituations1which,NRC wants to be notified prior to returning the S/Gs to service, For SR 4.4.5.5.d.,
Items 3 and 4, indications are applicable ~
only where alternate plugging criteria is being applied.
For the purposes of this'repcrting requirement,' leakage and conditional burst probability can be l
calculated based on the as-tound voltage distribution rather than the projected
~
. hen end-of-cycle voltage distribution (refer to GL 95-05 for more information) w it is not practical'to' complete these calculations using the projected.EOC voltage distributions prior to returning the S/Gs to service.. Note that.if leakage and conditional burst probability were calculated using the measured t
l EOC voltage distribution for the purposes of addressing GL Sections 6.a.1 and.
6 a.3 reporting criteria, then the results of.the projected EOC voltage distribution should be provided per GL Section 6.b(c) criteria.
l i
,t Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it 4
will be found during scheduled inservice steam generator. tube examinations.
Plugging will be -required for all tubes with imperfections exceeding the repair i
limit defined in Surveillance Requirement 4.4.5.4.a.
The portion of the tube that the plugging limit does not apply to is the portion of-the tube that is not.within the RCS pressure boundary (tube end up to the start of. the tube-to-R193 tubesheet weld). The tube end to tube-to-tubesheet weld portion of the tube 2.
does not affect structural integrity of the steam generator. tubes and therefore indications found in this portion of the tube will be excluded from the Result and Action Required for tube inspections.
It is expected that any indications that extend from this region will be detected during the scheduled tube inspections.
Steam generator tube inspections of operating plants have demonstrated'the capability to reliablv detect degradation that has penetrated 20% of the original tube wall thickness.
t I
Tubes experiencing outside diameter stress corrosion cracking within the thickness of the tube support plate are plugged or repaired by the criteria of 4.4.5.4.a.10.
i,
'Whenever the results of any steam generator tubing inservice inspection i
fall into Category C-3, these results will be promptly reported to the i
Commission pursuant to Specification 6.6.1 prior to resumption of plant opera-R40 tion.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests,
- additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
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SEQUOYAH - UNIT 1 B 3/4 4-4 Amendment No. 36, 189, 214,222 j
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' 3 / 4. 4 '. 6 REACTOR COOLANT SYSTEM LEAKAGE
, o 3/4.4.6.1' LEAKAGE DETECTION SYSTEMS I
The RCS leakage detection systems required by this specification are.
. pre-ided to monitor and detect leakage from the Reactor Coolant Pressure Boundary., These detection systems are consistent with the recommendations of
-Regulatory Guide 1,45,1" Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
J 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 surveillance requirements for RCS Pressure Isolation Valves provide added assurances of valve integrity thereb'y reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS-isolation valves is= IDENTIFIED LEAKAGE and will be considered as a portion of the allowed l-limit.
The 10'GPM. IDENTIFIED LEAKAGE limitation provides-allowance for a limited.
amount of leakage from known. sources whose presence will not interfere with the 1-
~ detection-of UNICENTIFIED LEAKAGE by the leakage detection sy.em.s.
~
~The CONTRCLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 GPM with the modulating i
valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
j This limitation ensures that in the event of a LOCA, the safety injection flow 4
will not be less than assumed in the accident analyses.
The total steam generator tube leakage limit of 600. gallons per day for-all steam generators and 150 gallons per day for any one steam generator will minimize the potential for a significant leakage event during steam line break-.
Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from.the previous inspection, the expected leak rate following a steam line rupture is limited to below 3.7 gpm in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 100 CFR :.00 guidelines.
If the projected and cycle distribution of crack indications results'in primary-to-secondary leakage greater than 3.7 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 3.7 gpm.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to
=
be promptly placed in COLD SHUTDOWN.
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.SEQUOYAH - UNIT 1 B 3/4 4-4a Amendment No. 36, 189, 214, 222 1
i 4
4
-4,-
p> >* "thf g~
-4 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION e
WASHINGTON. D.C. M56H001 o
- % *..,
- yd TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SE0V0YAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 213 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated October 18, 1996 as supplemented March 12, March 17, April 4, and April 9, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
i and the Commission's rules and regulations set forth in 10 CFR i
Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and i
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations, i
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and
]
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
i 1
J 2.
Accordingly,. the license is amended by chtnges to the Technical Specifications as indicated in the attachment to this license amendment.
Paragraphs 2.C.(2) and 2.C.(9) of facility Operating License No. DPR-79 are hereby amended to read as follows:
(2) Technical Sqqcifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.213, are hereby incorporated in the license.
The 1.censee shall operate the facility in accordance with the Technical Specifications.
A new condition 2.C.(8)(b) is added as follows:*
i (9) Steam Generator Insoection (Section 5.3.1)
~
(b)
By May 20, 1997, TVA shall establish a steam generator
. inspection program that is in accordance with the commitments listed in Enclosure 2 to the TVA letter to the Commission on this subject dated March 12, 1997, as modified by TVA letter dated March 17, 1997.
3.
This license amendment is effective as of its date of issuance, to be implemented no later than 45 days of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
i.s L.b b,
Frederick J. Hebdon, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
1.
Page 4 of License 4
2.
Changes to the Technical Specifications j
Date of Issuance: April 9, 1997 3 Page 4 of the composite license is attached to reflect this change i
d.
Failure to complete any tests included in the described program t
(planned or scheduled) for power levels up to the authorized power level.
(4)
Monitorine Settlement Markers (SER/SSER Section 2.6.3)
TVA shall continue to monitor the settlement markers along the EPCW conduit for the new ERCW intake structure for a period not less than three years from the date of this license. Any settlement greater than 0.5 inches that occurs during this period will be evaluated by TVA and a report on this matter will be submitted to the NRC.
(5)
Tornado Missiles (Section 3.5)
Prior to startup after the first refueling of the facility, TVA shall reconfirm to the satisfaction of the NRC that adequate tornado protection is provided for the 480 V transformer ventilation systems.
(6)
Qcigaff Seismie Catecorv Structures (Section 3.8)
Prior to startup following the first refueling, TVA shall evaluate all seismic Category I masonry walls to Saal NRC criteria and implement NRC required modifications that are indicated by that evaluation.
(7)
Low Temperature Overpressure Protsglism (Section 5.2.2)
Prior to startup after the first refueling, TVA shall install an overpressure mitigation system which meets NRC requirements.
(8)
Steam Generator Insnectica (Section 5.3.1)
(a)
Prior to start-up after the first refueling, TVA shall install inspection ports in each steam generator or have an alternative for. inspection that is acceptable to the NRC.
(b)
By May 20,1997, TVA shall establish a steam generator inspection program that is in accordance with the commitments listed in to the TVA letter to the Commission on this subject dated March 12,1997, as modified by TVA letter dated March 17, 1997.
(9)
Containmer t Isolation hstems (Section 6.2.4)
Prior to startup after the first refueling, TVA shall modify to the satisfaction of the NRC the onc-inch chemical feed lines to the main and auxiliary feedwater lines for compliance with GDC 57.
(10)
Environmental Oualification (Section 7.2.2j a.
No later than June 30,1982, TVA shall be in compliance with the requirements of NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,"
for safety-related equipment exposed to a harsh environment.
Amendment No. 213
ATTACHMENT TO LICENSE AMENDMENT NO. 213 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET N0, 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and ir,serting the enclosed pages.
The revised pages are identified by the cap'.ioned amendment number and contain marginal lines indicating the area cf change.
RM2Y1 INSERT 3/4 4-11 3/4 4-11 3/4 a-13 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-14a 3/4 4-14a 3/4 4-14b 3/4 4-14b 3/4 4-18 3/4 4-18 8 3/4 4-3 8 3/4 4-3 8 3/4 4-3a B 3/4 4-3a B 3/4 4-4 8 3/4 4-4 e
2 I
r e
P
F REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1.
All nonplugged tubes that previously had detectable wall pene-trations (greater than 20%).
2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to specification 4.4.5.4.a.0) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
4.
Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.
c.
The tubes selected as the second and third samples (if required by Table 4,4-2) during each inservice inspectica may be subjected to a partial tube inspection provided:-
1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
j 2.
The inspections include those portions of the tubes where l
imperfections were previously found.
Note:
Tube degradation identified in the portion of the tube that R181 i
is not a reactor coolant pressure boundary (tube end up to the start of the tube-to-tubesheet weld) is excluded frem l
the Pesult and Action Required in Table 4.4-2.
I d.
Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest i
cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indicitions. The determination of the l
lowest cold-leg 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.
The results of each sample inspection shall be classified into one of the following three categories:
Catenerv Inspection Results C-1 Less than 5% of the total tubes inspected are 1
degraded tubes and none of the inspected tubes are defective.
1 SEQUOYAH - Unit 2 3/4 4-11 Amendment No. 181, 211, 213
e REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceotance Criteria a.
As used in this Specification:
1.
Imoerfection 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 2C% of the nominal tube wall thickness, if detectable, may be con-sidered as imperfections.
2.
Decradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
3.
R2craded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.
t Deoradation means the percentage of the tube wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defectivc.
6.
Pluccine Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.
Plugging limit does not apply to that portion of the tube that is not within the pressure R181 boundary of the reactor coolant system (tube end up to the start of the tube-to-tubesheet seld).
This definition does not apply to tube support plate intersections if the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.
i 7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural 1
integrity in the event of an Operating Basis Earthquake, a loss-j of-coolant accident, or a steam line or feedwater line break as 4
specified in 4.4.5.3.c, above.
i 8
Igbe Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
9.
Preservice Insoection means an inspection of the full length of each tube in each ste?m generator performed by eddy current techniques prior to service to establish a baseline condition of i
the tubing. This inepection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10.
Tube Succort Plate Pluccinc Limil is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of the i
SEQUOYAH - UNIT 2 3/4 4-13 Amendment No. 181, 211,213
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1 9
,1
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A j
SURVEILLANCE REQUIREMENTS (Continued) j At tube support plate intersections,1the i
tube support plates._
O plugging (repair) limit is based on maintaining steam generator.
i
. tube serviceability as, described below V
a.
. Steam generator tubes,-whose degradation is attributed to.
outside diameter stress corrosion cracking within the-bounds of the tube *apport plate with bobbin. voltages less o
than or equal'to the lower vtitage repair limit (Note 1),
I will:be' allowed to remain.in service.
r b.
Steam generator tubes, whose degradation is attributed:to
.outside diameter stress corrosion cracking within the 3
bounds of.the tube support plate with a bobbin voltage greater.than the lower voltage repair limit-(Note 1), will 4
.be repaired or plugged, except.as noted in 4.4.5.4.a.10.c below.
c.
Steam generator tubes, with indications of' potential =
E degradation attributed to outside diameter stress-l-
. corrosion-cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage 3
repair limit (Note 1), but less than or equal to upper voltage repair limit (Note 2), may remain in service if a retating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of j
outside diameter stress corrosion-cracking degradation with a bobbin coil voltage greater than the upper voltage repair limit (Note 2) will be plugged or repaired.
d.
Not applicable to SQN.
e.
If an unscheduled mid-cycle inspaction is performed, the follow'ng mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.
The mid-cycle repair limits are determined from the following equations:
V" Va=
- 1. 0 + NDE + Gr CL Va=Va - (Vm - Vzu)
} s t)
SEQUOYAH.-l UNIT 2 3/4 4-14 Amendment No. 28, 211,213 l
i
o EEACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) where:
upper voltage repair limit Vunt
=
lower voltage repair limit vtat
=
uunt mid-cycle upper voltage repair limit based on time v
into cycle i
ut,t mid-cycle lower voltage repair limit based on Vgat and V
=
time into cycle length of time since last scheduled inspection during At
=
which v and Vat were implemented uat t
cycle length (the time between two scheduled steam CL
=
generator inspections) structural limit voltage Vst
=
average growth rate per cycle length Gr
=
95-percent cumulative probability al3owance for NDE
=
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.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.
Note 1:
The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing or 2.0 volts for 7/A-inch diameter tubing.
Note 2:
The upper voltage repair limit is calculatad acccrding to the methodology in GL 90-05 as supplemented. Vcat may differ at the TSPs and flow distribution baffle.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Recortq a.
Following each inservice inspection of steam generator tubes, the number of tubes plugged 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 submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.
This Special Report shall include:
1.
Number and extent of tubes inspected.
SEQUOYAH - UNIT 2 3/4 4-14a Amendment No. 28, 211,213
~...
le E
. e EEACTOR COOLANT SYSTEM
' SURVEILLANCE! REQUIREMENTS (Centinued).
2.
Location and' percent of wall-thickness' penetration for each indication of an imperfeccion.
3.
Identification lof. tubes plugged.
l 4
c.
Results of steam generator tube inspections which' fall into' Category R28~
C-3 shall be reported pursuant to Specification 6.6.1. prior to
= resumption of plant operation. The written followup of this report shall provide a description of investigations conducted.to determine 3:
- cause of the tube degradation and corrective measures taken to prevent recurrence.
I d.
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returning-the i
steam generators to service should any of the.following conditions-arises (or -: if j
1.
If. estimated leakage based on the projected end-of-cycle 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 steam line break) for the next operating cycle.
1 l
2.
' If circumferential crack-like indications are detected at the tube support plate intersections.
3.
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.
5.
If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual
]
i measured end-of-cycle) voltage distribution exceeds 1 X 103, notify the NRC and provide an assessment of the safety significance of the occurrence.
\\
l i
i SEQUOYAH
" UNIT 2 3/4 4-14b Amendment No. 28, 211, 213 r.
4
-e.
. ru 4
o.-
.. ~. _
^'
b f
y REACTOR COOLANT SYSTEM OPERATIQtTM, ~ LEAKAGE t
. LIMITING CONDITION-FOR OPERATION
.3.4.6.2' Reactor' Coolant' System leakage shall be limited to:
a.
-b.
-1 GPM UNIDENTIFIED LEAKAGE,
- j 150 gall'ons per day of primary-to-secondary leakage through any one c.
steam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,'and e.
40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of' 2235 1 20 psig.
~
f.
1 GPM leakage at a Reactor Coolant System pressure of 2235 i 20'psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
3 j
APPLICABILITY: MCTES 1, 2, 3 and 4 ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within l
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 ::,ne of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 nours 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 />.
i 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 closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
[
SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within l-
.each of the above limits by:
f
-i I
-SEQUOYAH - UNIT'2 3/4 4-18 Amendment No. 211,213 i
t
e e
REACTOR COOLANT SYSTEM BASES 314.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is
'hased 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'eause 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 coola.it 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 (prima ry-to-secondary leakage = 150 gallons per day per steam generator).
Cracks having l
a primary-to-secondary leakage less than this limit during operatien will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Sequsyah has demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown or condenser off-gas.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
i The voltage-based repair limits of SR 4.4.5 implement the guidance in GL 95-05 and are applicable only to Westinghouse-designed steam generators (S/Gs) with outside diameter stress corrosion cracking (ODSCC) Iccated at the 1
tube-to-tube support plate intersections. The voltage-based repair limits are not applicable to other forras of S/G tube degradation nor are they applicable j
j to ODSCC that occurs at other locations within the S/G.
Additionally, the
)
repair criteria apply only to indications wherr. the degradation mechanism is I
dominantly axial ODSCC with no significant cracks extending outside the thickness of the support plate.
Refer to GL 95-05 for additional description of the degradation morphology.
Implementation of SR 4.4.5 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 flaw growth during an operating interval and to account for NDE uncertainty. The upper voltage repair limit; Vat, is determined from the structural voltage limit by cpplying the following equation:
V aL " Vst - Vca -V u
.SEQUOYAH - UNIT 2 B 3/4 4-3 Amendment No. 181, 211,213
m
~
0'
'e IREACTOR COOLANY SYSTEM BASES!
where Vca represents the allowance for flaw growth between inspections and'v g e
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 of SR 4.4.5.4.a.10.e should only be used during unplanned inspection in which eddy current data is acquired for indications at i
the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommended by
~
GL 95-05 for situations which NRC wants to be notified prior to returning the S/Gs'to service.
For'SR 4.4.5.5.d.,-Items 3 and 4, indications are applicable t
only where alternate plugging criteria is being applied.
For the purposes of this reparting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when Lit,is not practical to complete these calculations using the projected EOC voltage. distributions prior to returning the S/Gs to service. Note that if leakage and conditional burst probability were calculated using the measured EOC. voltage distribution for the purposes of addressing GL Sections 6.a.1 and l
6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per GL Section 6.b(c) criteria.
-Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant, However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the repair limit defined in Surveillance Requirement 4.4.5.4.a.
The portion of the tube that the plugging limit does not apply to is the portion of the tube that is not within the RCS pressure coundary (tube end up to the start of the tube-to-R181 tubesheet weld).
The tube end to tube-to-tubesheet weld portion of the tube does not affect structural integrity of the steam generator tubes and therefore indications found in this portion of the tube will be excluded from the Result and Action Required for tube inspections.
It is expected that any indications that extend from this region will be detected during the scheduled tube inspections.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
l Tubes experiencing outside diameter stress corrosion cracking within the thickness of the tube support plate are plugged or repaired by the criteria of 4.4.5.4.a.10.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, chese results will be promptly reported to the Commission pursuant to Specification 6.6.1 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, inboratory examinations, tests, additional eddy-current inspection, and revision of.the Technical Specifications, if necessary, i'
SEQUOYAH - UNIT 2 B 3/4 4-3a Amendment No. 181, 211,213 n
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6.
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-e REACTOR COOLANT SYSTEM' BASES
~ 3/4.4.6 REACTOR COOLANT SYSTEM LEAFAGE jfju4.6.1 LEAFAGE 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 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 additienal leakage.
The surveillance requirements for RCS Pressure Isolation Valves provide added assurances of valve integrity thereby reducing the probability of gross-valve failure a..d consequent intersystem LOCA.
Leakage from the RCE isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.
The total steam generator tube leakage limit of 600 gallons per day for all steam generators and 150 gallons per day for any one steam generator will minimize the potential for a significant leakage event during steam line break.
Baspd on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 3.7 gpm in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines.
If the projected and cycle distribution of crack indications results in primary-to-secondary leakage greater than 3.7 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed f rom service in order to reduce the postulated primary-to-secondary steam line break leakage to below 3.7 gpm.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross f ailure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
'SEQUOYAH - UNIT 2 B 3/4 4-4 Amendment No. 211,213