ML20129D266
| ML20129D266 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/18/1996 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20129D262 | List: |
| References | |
| NUDOCS 9610240248 | |
| Download: ML20129D266 (41) | |
Text
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s 1,
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i ENCLOSURE 1 i
PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE i
SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS 50-327 AND 50-328 (TVA-SON-TS-96-05)
)
l l
i LIST OF AFFECTED PAGES i
Unit 1 1
3/4 4-9 3/4 4-9a 3/4 4-9b 3/4 4-10 3/4 4-14 1
B 3/4 4-3 B 3/4 4-4
}
B 3/4 4-4a t
Unit 2 l
3/4 4-11 4
3/4 4-13 3/4414 3/4 4-14a 3/4 4-14b 3/4 4-18 f
B 3/4 4-3 B 3/4 4-3a B 3/4 4-4 i
9610240248 961018 PDR ADOCK 05000327 P
1 4
s
_ REACTOR COOLANT SYSTEM SURV U LLANCE REQUIREMENTS (Continued) l 4.4.5,4 Accentance Criteria As used in this Specification:
l a.
Inndefectionmeansanexceptiontothedimensions, finisher
(
1.
contour of a tube from that required by fabrication drawings l
specifications.the nominal tube wall thickness, if detectable, may be con-i i
l sidered as imperfections.
l
Dearadation means a service-induced cracking,
wastage, wear or l
general corrosion occuring or either inside or outside of a tube.
2.
Dearaded Tube means a tube containing imperfections greater than or equal to.20% of the nominal wall thickness caused by 3.
j degradation,
[
5 Deoradation means the percentage of the tube wall thicliness f
4.
affected or removed by degradation.
Dgiggi means an imperfection of such severity that it exceeds the A tube containing a defect is defective.
S.
l plugging limit.
j Pluacina limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of t 6.
l R193 nominal tube wall thickness.
pressure boundary i
that portion of the tube that is not within thmtube end up to the start of t i
i F
of the reactor coolant system (definition does not apply to tube g
This tube-to-tubasheet weld).
support plate intersections if the voltage-based repair criteria i
R218 J
Refer to 4.4.5.4.a.10 for the repair limit 4
are beine applied.
applicable to these intersections.
l Unserviceable describes the condition of a tube if it leaks contains a defect large enough to affect its structural.
7.
integrity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line i
l break as specified in 4.4.5.3.c, above.
Tube intnection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the 8.
U-band to the top support of the cold leg.
Preservice Insnection means n' tube inspection of the full Tength of each tube in each steam generator prformed by e 9.
current techniques prior to service establism a baseline co dition of the tubing.
to initial POWER OPERATION using the equipment and technique expected to be used during subsequent inservice inspections.
Tube Sunnert Plate Pluccina Limit is used for the disposition 18 of an alloy 600 steam generator tube for c 10.
stress corrosion cracking confined within the Scat d ch::;e: te tMc p:; are-app 14 cable-to-Cycle-s-operati::: ::1
- Tee 4:
/
Amendment No. 189, 214 3/449 SEQUOYAH - UNIT 1 October 11. 1999
s REACTOR COOLANT SYSTEM SURVEILLkNCEREQUIREMENTSIContinued) i thickness of the tube support plates. At tube support plate g3 X1 inter' ections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the a.
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.
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the b.
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 below.
Steam generator tubes, with indications of potential c.
degradation attributed to outside' diameter stresscorros plate with a bobbin voltage greater than the lower voltage i
repair limit (Note 1), but less than or equal to upper, may rem voltage repair limit (Note 2) ion does not detect rotating pancake coil inspectSteam generator tubes, with indication degradation.outside diameter stress corrosion-cracking degradation with a bobbin coil voltage greater than the upper voltage repair limit (Note 2) will be plugged or repatrod.
d.
Not applicable to SQN.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead e.
The mid-cycle repair limits are determined from the following equations:
Y, g,n g g 4 (CL-At)
CL Ya=Vm-(Y g-YQ(CL-AQ a
f-indiceUd cher,g;; tMhis p;g; ere epplicable t; Oycle
- epsretier, er.174
.;:a
- Th:
~
October 11, 1995 Amendment No. 189, 214 3/4 4-9a -
SEQUOYAH - UNIT 1
1 i
PEACTOR COOLANT SYSTEM 4
l ;
SURVEILLANCE REOUIREMENTS (Continued 1 f
where:
218 upper voltage repair limit l
V.
lower voltage repair limit V
m mid-cycle upper voltage repair limit based on time
)
j V.
into cycle i
mid cycle lower voltage repair limit based on Vgg Q and time into cycle i
length of time since last scheduled inspection At
=
were implemented during which V. and Vm cycle length (the time between two scheduled steam ct
=
generatorinspections) structural limit voltage V
a average growth rate per cycle length I
Gr I
g5-percent cumulative probability allowance for NDE l
nondestructive examination uncertainty (i.e., a i
value of 20-percent has been approved by NRC)
~
l' Implementation of these mid-cycle repair limits should follow the same f}-
approach as in TS 4.4.5.4.a.10.a. 4.4.5.4,s.10.b, and 4.4.5.4.a.10.c.
l The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing Note 1:
or 2.0 volts for 7/8-inch diameter tubing, l
The upper voltage repair limit is' calculated according to the methodology in GL 95-05 as supplemented. V may differ at ;he TSPs and flow j
Note 2:
m 4
distribution baffle.
i i
i i
i s
l 2
1 4dicated-ch;ng:: te-tMs ;;;; Orc :pplic;51; to Cy;le 0 ;pr;t!;; = 7
- Th:
ivnendment No.18g 214 3/4 4-9b SEQUOYNd - UNIT 1 October 11, 1995 4
-__m_..
l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continuedi The steam generator shall be determined OPERABLE after coepleting f
b.
the corresponding actionf. (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) r6 quired by Table 4.4-2.
4.4.5.5 Reoorts Fo110wird each inservics inspection of steam generator tubes, the number of tubes,nluc;ad in each steam generator shall be reported a.
to the Consission within 15 days.
The complete results of the steam get.e. tor tube inservice b.
inspection shall be su%sittd to the Counission in a Special Report pursuant to Specification 6.g.2 within 12 months followingThis Special completion of the inspection.
1.
Number and extant of tubes inspected.
Location and percent at wall-thickness penetration for each 2.
indication of.an imperfection.
3.
Identification of tubes plugged, Result's of steam generator tube inspections whien fall into R40 Category C-3 shall be reported pursuant to Specification 6 6.1The wr c.
prior to resumption of plant operation.this report shall provid to determine cause of the tube degradation ard corrective measures taken to prevent recurrence.
For implementation of the voltage-based repair criteria to tube R218 support plate intersections, notify the sta# nrior to returning d.
the steam generators to service should any.of the following conditions arise:
If estimated leakage based on the projected end-of-cycle (or if not practical using the actual measured end-of-1.
cycle) voltage distribution excueds the leak limit (determined from the licensing tasis do:e calculation for the postulated main steam line b.eak) for the.M xt operating cycle.
If circumferential crack-like indications are dete:ted at 2.
the tube support plate intersections.
If indications are identified that extand beyond the 3.
confines of the tube support plate.
If indications are identified at the tube support plate 4.
elevations that are attributablesto primary water stress corrosion cracking.
If the calculated conditional burst probability based on f
the projected end-of-cycle (or if not practical, using the 5.
actual pasured end-of-cycle) voltage distribution exceed i
1 x 10', notify the NRC and provide an assessment of the safety significance of the occurrence.
- ne-tediest4ch;a;;cs to thh p:;c rc :pplicebh-to-Gyeh-o-operation-enly2fp-Amendment No. 36, 3/44-10 214 SEQUOYAH - UNIT 1 October 11. 1995
-=
i l
a I
,i,
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION
)
Reactor Coolant System leakage shall be limited to:
3.4.6.2 No PRESSURE BOUNDARY LEAKAGE, a.
l b.
1 GPM UNIDENTIFIED LEAKAGE,
^1 OPM-total-primary-to-secondary-leakage-through-all-steam-senerators--
end 500-gallens per day-through-4ny ::: :tean-genereten
-c.
Q218
$.150gallonsperdayofprimary-to-secondaryleakagethroughany l
uteam generator, 10 GPM IDENTIFIED LEAKAGE from th's Reactor Coolant System, and I
d.
'40 GPM CONTROLLED LEAXAGE at a Reketor Coolant System e.
pressure of 2235 t 20 psig.
2235 1 20 psig l
1 GPM leakage at a Reactor Coolant System pressure of R16 from any Reactor Coolant System Pressure Isolation Valve specified i f.
l i
Table 3.4-1.
J APPLICABILITY: MODES 1, 2, 3 and 4 f
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be a.
With any Reactor Coolant System leakage gr$ater than any one of t above limits, excluding PRESSURE BOUNDARY b.
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 $HUTDOW
.j'
?
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With any Reactor' Coolant System Pressure Isolation Valve c.
greater than the above limit, 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 the affected system from the use of at least two closed manual or deactivated automatic valves or be in at least NOT STANDBY within the next 6 ho l
SHUTDOWN within,the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l SURVEILLANCE REQUIREMENTS R16 f
Reactor Coolant System leakages shall be demonstrated to be 4.4.6.2.1 kig each of the above limits by:
^
- RepMcement-of-crwith-cc,--is-appl 4 sable-for-Gysle-8-eperat4cn
~
l CE Amendment No. 12, 214 c v 3/4 4-14 SEQUOYAH - UNIT 1 october 11, 1995
l l
REACTOR COOLANT SYSTEM BASES The plant is expected to be operated in a manner such that the secondary coolant will be. maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during
~
plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 150 gallons per day per steam generator).
h218 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 per steam generator 218 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 Itcated and plugged.
The voltage-based repair limits of SR 4.4.5 implement the guidance in GL hig 95-05 and are applicable only to Westinghouse-designed steam generators (S/Gs) with outsit.e diameter stress corrosion cracking (00 SCC) located at the tube-to-tube support plate intersections. The voltage-based repair limits are not applicable to other foms of 5/G tube degradation nor are they applicable to 00 SCC that occurs at other locations within the S/G. Additionally, the repair criteria apply only to indications where the deg*adation mechanism is dominantly axial 00$CC with no significant cracks extending outside the thickness of the support pitte. Refer to GL 95-05 for additional description of the degradatfon morphology.
J Implementation of SR 4.4.5 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the 3
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 g5 percent prediction interval curve reduced to account for the lower 95/95 percent tolerance bound for tubine material i.e., the 95 percent LTL curve). The voftage structural properties at 650*F (d downward to account for potential flaw growth during an limit must be adjuste operating interval and to account for NDE uncertainty. The upper voltage repair limit; V is determined from the structural voltage limit by applying the following e, ion:
quat V
= Va - V - V.,
represents the allowance for flaw growth between inspections and V where V"nts the allowance for potential sources of error in the measurement oT represe the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL g5-05.
Orc :pheeble Nych a hration-eMr tu che iies to-tMs p:;:
7 rh SEQUOYAH - UNIT 1 B 3/4 4-3 Amendment No. 36, 18% 214 October 11, 1995
l
}
RFACTOR COOLANT SYSTEM BASES The mid-cycle equation of SR 4.4.5.4.a.10.e should only be used during
@ig f
unplanned inspection in which addy current data is acquired for indications at l
the tube support plates.
}
SR 4.4.5.5 implements several reporting requirements recommended by GL g5-05 for situations which NRC wants to be notified prior to returning the S/Gs to 4
For SR 4.4.5.5.d Items 3 and 4, indications are applicable only i
For the purposes of this service.
where alternate plugging criteria is being applied.
i reporting requirement, leakage and conditional burst probability 1
j end-of-cycle voltage distribution (refer to GL g5-05 for more information) when it is not practical to complete these calculations using the projected E0C l
Note that if voltage distributions prior to returning the S/Gs to service. leakage l
EOC voltage distribution for tie purposes of addressing GL Sections 6.t.1 and 6.a.3 reporting criteria, then the results of the projected E0C voltage distribution should be provided per GL Section 6.b(c) criteria, l
i Wastage-t 1 defects are unitkely with proper chemistry treatme f
However even if a defect should develop in service, it secondary cool n.
will be found during scheduisd inservice steam generator tube examinations.
j Plugging will be required for all tubes with imperfections exceeding the repair
%218 11mit defined in Surveillance Requirement 4.4.5.4.a.
The portion of the tube l
that the plugging limit does not apply to is the portion of the tube that is l
not within the RC5 pressure boundary (tube end up to the start of the tube-to-i R193 The tube and to tube-to-tubesheet weld portion of the tube tubesheet weld).
does not affect structural integrity of the steam generator tubes and therefore l
indications found in this portion of the tube will be excluded from the Result j
and Action Required for tube inspections. It is expected that any indications that extend from this region will be detected during the scheduled tube Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated inspections.
20% of the original tube wall thickness.
219 l
Tubes experiencing outside diameter stress corrosion cracking within the thickness of the tube support plate are plugged or repaired by the criteria of j
4.4.5.4.a.10.
j Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.6.1 prior to resumpt R40 tion.
- tests, and may result in a requirement for analysis, laboratory examinations,ifications, additional eddy-current inspection, and revision of the Technical Spec if necessary.
hT.% ;Micate chc;;c; te this ;;;: cre c;;1hdic te Cycle 0 eHietiG 1l s@
Amendment No. 36, 18g, 214 "
8 3/4 4-4 SEQUOYAH - UNIT 1 October 11, 1995
l
+
l i
BASE 5 l
3/4.4.6 _ REACTOR C00LANT,,1YSTEM 4AKAGE 3/4.4.6.1 LEAKAGE DETECTION $11I[ $
The RCS leakaSe 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 uduced 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 thereby 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 limit.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited A
amount of leakage from known sources whose presence will not interfere with the G
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 modulatin valve in the supply line fully open at a nominal RCS pressure of 2235 psig. g 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
$18 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 gpa in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines. If the projected end of cycle distribution of crack indications results in primary d to-secondary leakage greater than 3.7 gpa in the faulted loop during a postulate steam line break event, additional tubes must be removed from servico in order to reduce the postulated primary-to-secondary steam line break leakage to below 3.7 gps.
PRESSURE B0UNDARY 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.
a tr4;cated-cherg;; to thh p;;; =0 :ppl4c:ble t0 Cycl: ! ;?r:t p218 SEQUOYAH - UNIT I B 3/4 4-4a Amendment No. 36, 189, 214 October 11, 1995
t J
~,) SURVEILLANCE REQUIREMENTS(Continued) l
'l 1.
All nonplugged tubes that previously had detectable wall pene-trations (greater than 20%).
]
2.
Tubes in those areas where experience has indicated potential i
problems.
3.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be j
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 R 11 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 inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found.
Note:
Tube degradation identified in the portion of the tube that R181 is not a reactor coolant pressure boundary (tube end up to the start of the tube-to-tubesheet weld) is excluded from the Result 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 R211 and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the 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:
Catecorv Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
3 nditated-changca to this-page-ere-appl-ieable-to-Gyele S operation ely.
R211 6%
April 3, 1996 SEQUOYAH - Unit 2 3/4 4-11 Amendment No. 181, 211
'N
'l SURVEILLANCE REQUIREMENTS (Continued)
/
4.4.5.4 Acceptance Criteria a.
As used in this Specification:
1.
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 con-sidered as imperfections.
2.
Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
3.
Decraded Tube means a tube containing imperfections greater than j
or equal to 20% of the nominal wall thickness caused by j
degradation.
j 4.
% Deoradatian means the percentage of the tube wall thickness affected or removed by degradation.
1 5.
Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.
6.
Pluccina 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 1
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 weld).
This definition does not apply
[8) to tube support plate intersections if the voltage-based repair R211 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 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.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
9.
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 prior to initial POWER OPERATION using the equipment and techniques expected to I
be used during subsequent inservice inspections.
10.
Tube Support Plate Pluccina Limit is used for the disposition of
)()
an alloy 600 steam generator tube for continued service that is R211 experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of the ndicated changcs to thic page arc applicable tc cycle e-operation _3, 1996 Apr11 l
SEQUOYAH - UNIT 2 3/4 4-13 Amendment No. 181, 211
~.. __
i j
-REACTOR COOLANT SYSTEM j
SURVEILLANCE REQUIREMENTS (Continued) t tube support plates. At tube support plate intersections, the f{}
'j plugging (repair) limit is based on maintaining steam generator h211 tube serviceability as described below:
Steam generator tubes, whose degradation is attributed to a.
i outside diameter stress corrosion cracking within the j
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 to j
outside diameter stress corrosion cracking within the 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 j
below.
1 1
Steam generator tubes, with indications of potential c.
j degradation attributed to outside diameter stress 1
corrosion-cracking within the bounds of the tube support l
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 j
rotating pancake coil inspection does not detect 1
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.
a i
d.
Not applicable to SON.
i If an unscheduled mid-cycle inspection is performed, the l
e.
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:
V
- 1. 0 + NDE + Gr ( CL-a t)
CL i
i
~ # E}
- (V
-Vm)
Va = Vm m
4 i
d h ndieeted chang:0 to thic p:30 Orc applicable t0 Cy010 S Op0rsti0n On YY April 3, 1996 SEQUOYAH - UNIT 2 3/4 4-14 Amendment No. 28, 211 A
)
w
)
SURVEILLANCE REQUIREMENTS (Continued) h where:
R211 upper voltage repair limit V nt u
lower voltage repair limit Vua
=
mid-cycle upper voltage repair limit based on time V unt M
into cycle mid-Cycle lower voltage repair limit based on V OUd MURL V utL
=
M time into cycle length of time since last scheduled inspection during At and Vat were implemented which Vunt 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 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
,7 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.
y.
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 methodology in GL 90-05 as supplemented. Vunt may dif fer 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 Reports Following each inservice inspection of steam generator tubes, the a.
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:
I 1.
Number and extent of tubes inspected.
'Thc indicated chcngca to thic pcg: crc cppMeabic to Cycle S cperation-onby-r<e April 3, 1996 Amendment No. 28, 211 SEQUOYAH - UNIT 2 3/4 4-14a l
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged.
Results of steam generator tube inspections which fall into Category R28 c.
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 h) support plate intersections, notify the staff prior to returning the R211 steam generators to service should any of the following conditions arise:
1.
If estimated leakage based on the projected end-of-cycle (or if not practical using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.
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 i
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 10'2, notify the NRC and provide an assessment of the safety significance of the occurrence.
I t-The-indicated-ehang 00 tO trh-ic POE0 OTO Op NIO3D10 UO CYO10 9 OECIII U U 211
.M.E t
April 3, 1996 i
SEQUOYAH - UNIT 2 3/4 4-14b Amendment N's.
28, 211
__.m..
_____.m___
m____..__
I r
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall'be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, c.
1 C= tet:1 prir r/-te ::::ndtry 10th:;: thr:u;F 11 :te: ; :: ret +ee--
- d 5^^ ;:ll::: per d y through any rre eter ;:::::t::,
h.
150 gallons per day of primary-to-secondary leakage through any one R211 steam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.
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 Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, beinatlhastHOTSTANDBYwithin
'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, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage j
rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With any Reactor Coolant System Pressure Isolation Valve leakage c.
greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two 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 each of the above limits by:
- Replse:- -t cf c.
- ith ec. 1 epplic-'le fer Cycle e eperatier enly.
13
.-i %
?m'C April 3, 1996 SEQUOYAH - UNIT 2 3/4 4-18 Amendment No.=211
o l
BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be 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-(k) to-secondary leakage = 150 gallons per day per steam generator).
Cracks having lR211 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-3)
)
to-secondary leakage of 150 gallons per day per st'eam generator can readily be R211 detected by radiation monitors of steam generator blowdown or condenser off-gas.
Leakage in excess of this limit will require plant shutdown and an eT unscheduled inspection, during which the leaking tubes will be located and
.J plugged.
The voltage-based repair limits of SR 4.4.5 implement the guidance in h)
GL 95-05 and are applicable only to Westinghouse-designed steam generators R211 i
(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 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 applying the following equation:
Vat = Vst -Vca - VNDE
-- he-indicat-ed-changes to this page are applicable te Cycle 9 cperationenl[
April 3,
1996 SEQUOYAH - UNIT 2 B 3/4 4-3 Amendment No. 181, 211
BASES
@)
where v represents the allowance for flaw growth between inspections and vmx a
represents the allowance for potential sources of error in the measurement of R211 the bobbin coil vo3tage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
1 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 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 j
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 reporting 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 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 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.
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, g
Plugging will be required for all tubes with imperfections exceeding the repair
'N limit defined in Surveillance Requirement 4.4.5.4.a.
The portion of the tube R21 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-R181 j
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 i
20% of the original tube wall thickness.
Tubes experiencing outside diameter stress corrosion cracking within the f) thickness of the tube support plate are plugged or repaired by the criteria of R211 4.4.5.4.a.10.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.6.1 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, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
indicated-changcc to this page arc appl +eabic to-Cycle e cperation only.
b, R2 3j ~,%
April 3, 1996 SEQUOYAH - UNIT 2 B 3/4 4-3a Amendment No. 181, 211
.)
BASES 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 OPERATIONAL LEAKAGE Industry experience has shown that while a limited snount 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 thereby 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 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 GC all steam generators and 150 gallons per day for any one steam generator will R211 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 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 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.
R211 44 hndicated-changes to this pag; cre applicabic to Cycic 8 Operat-icn Only.
April 3, 1996 SEQUOYAH - UNIT 2 B 3/4 4-4 Amendment No. 211
pt
,,3-._.
4 4
ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-96-05)
DESCRIPTION AND JUSTIFICATION FOR TS AMENDMENT
4 Descriotion of Chanae TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 technical specifications (TSs) to remove cycle specific limitations associated with alternate steam generator (S/G) tube plugging criteria. SON TS 3/4.4.5, " Steam Generators,"
and the associated bases contain a footnote that states: "The indicated changes to this page are applicable to Cycle 8 operation only." TVA's proposed change removes I
this footnote from TS 3/4.4.5.
in addition, TS 3.4.6.2, " Operational Leakage," contains a similar footnote that states:
l
" Replacement of c. with cc. is applicable for Cycle 8 operation only." TVA's proposed change removes this footnote from TS 3.4.6.2 and deletes TS Action l
Requirement 3.4.6.2.c.
I Reason for Chanae TVA is proposing to change SON TSs to remove existing footnotes that limit the application of alternate tube plugging criteria (APC) to Cycle 8 operation only. The j
Cycle 8 limitation was applied to SON in 1995 pending resolution of two industry i
issues associated with probe wear and probe variability. These issues have now been resolved as documented in Nuclear Energy Institute (NEI) letters to NRC dated January 23 and February 23,1996. Accordingly, the Cycle 8 limitation is no longer i
applicable to SON, and the proposed TS change is needed to allow application of APC beyond Cycle 8 operation.
Justification for Chanaes The application of APC to SON S/Gs was implemented under TS Change 95-15 for Unit 1 and TS Change 95-23 for Unit 2.
in the safety evaluation report (SER) for TS Change 95-23,it was noted that probe wear and probe variability were the issues that remained to be resolved before APC could be applied to SON S/Gs on a permanent basis. Specifically,in Section 2.3 of the SER (Reference NRC letter to TVA dated April 3,1996), a discussion of the NRC staff position regarding probe wear and probe variability was provided. The following
{
is an excerpt from the SER:
i "Since issuance of GL 95-05, the staff has been working with industry through the Nuclear Energy Institute (NEI) on the issues of probe wear and new probe variability as they relate to sections 3.c.2 and 3.c.3 of Attachment 1 to GL 95-05. By two letters 4
dated January 23,1996, NEl proposed an alternative to the probe wear criteria in GL 95-05 and a methodology for implementing the 10 percent probe variability criteria. The staff approved the NEl's proposals subject to certain observations and restrictions in a letter from Brian Sheron of NRC to Alex Marion of NEl, dated February 9,1996. By a letter dated February 23,1996, NEl addressed the staff's observations and restrictions and agreed to supply certain confirmatory information. The licensee, 4
d
. in its letter dated March 4,1996, committed to implement the alternative criteria on probe wear and the industry methodology for limiting new probe variability as defined in the NEl's letters dated January 23 and February 23,1996. The licensee will ensure that the confirmatory information related to new probe variability has been provided to the staff prior to requesting permanent alternate plugging criteria for Sequoyah Units 1 and 2. The staff finds that the licensee's commitment is acceptable."
TVA hes included the recommendations and methods from the NEl letters referenced above into SON's S/G Program, in addition, confirmatory information requested in the SER has been provided to NRC in a NEl letter dated October 15,1996 ( letter from Alex Marion to Dr. Bryon W. Sheron entitled " Response to NRC Letter dated February 9,1996, Regarding New Probe Variability Criteria"). Accordingly, based on the recent resolution of these industry issues and the submittal of confirmatory information to NRC, the actions needed to support TVA's amendment request have been completed for permanent application of APC to SON Units 1 and 2.
Environmental Imoact Evaluation The proposed change does not involve an unreviewed environmental question because -
operation of SON Units 1 and 2 in accordance with this change would not:
1.
Result in a significant increase in any adverse environmentalimpact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.
2.
Result in a significant change in effluents or power levels.
3.
Result in matters not previously reviewed in the licensing basis for SON that may have a significant environmentalimpact.
i i
I I
ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-96-05)
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION 1
~
Significant Hazards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a sigreificant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not:
)
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
]
The proposed TS change revises SON steam generator (S/G) Specification 3/4.4.5 to remove footnotes that limit the application of the alternate plugging criteria (APC) to Cycle 8 operation only. In addition, SON TS 3.4.6.2, " Operational Leakage," contains a similar footnote that limits application of S/G APC to Cycle 8 operation only. The removal of these footnotes allows TVA to apply APC to SON S/Gs beyond Cycle 8 operation. TVA's proposed change is based on resolution i
of the industry issues concerning probe wear and probe variability. APC was
)
applied to the SON S/Gs during the Cycle 7 refueling outages for Units 1 and 2.
The proposed changes provide TS requirements that are consistent with the guidance of NRC GL 95-05. This change does not involve a physical modification to the plant or affect any setpoints. Accordingly, the proposed changes do not i
involve an increase in the probability or consequences of an accident previously evaluated.
2.
Create the possibility of a new or different kind of accident from any previously analyzed.
The proposed changes provide TS requirements for SON S/Gs that are consistent with the guidance provided in GL 95-05. No new event initiator has been created, nor has any hardware been changed. This change does not involve a physical change to SON S/Gs or any other system. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously analyzed.
3.
Involve a significant reduction in a margin of safety.
TVA's proposed change allows application of APC for SON S/Gs to extend beyond Cycle 8 operation. This change continues to provide requirements that maintain structuralintegrity of SON S/G tubes during normal operating, transient, and postulated accident conditions. This change does not involve e setpoint change or physical modification to the plant. Accordingly, the margin of safety has not been reduced.
=
=
s e
a a
au
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a da-..
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es.
e n+.m.
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.g t
ENCLOSURE 4 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-96-05)
REVISED TS PAGES l
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Accentance Criteria a.
As used in this Specifications 1.
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 J
be considered as imperfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occuring or either inside or outside of a tube.
3.
Decraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.
% Decradation means the percentage of the tube wall thickness affected or removed by degradation.
5.
ppfect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.
6.
Pluccing 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 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 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.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
9.
Preservice Inspection means a tube inspection of the full length of each tube in each steam generator' performed 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.
t 1
I 10.
Tube Suonort Platy Plucaina Limit 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 SEQUOYAH - UNIT 1 3/4 4-9 Amendment No. 189, 214,
i t
l l
j
~
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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 l
below a.
Steam generator tubes, whose degradation is attributed
)
l to outside diameter stress corrosion cracking within l
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 to outside diameter stress corrosion cracking within the 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 below, c.
Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion-cracking within the bounds of the tube support plate with a bobbin voltage greater than 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 does not detect degradation.
Steam generator tubes, with indications of 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 SON.
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:
V uRL
=
M 1.0+NDE+Gr CL (CL-At)
VMLRL=VMURL~ (YURL-V at) t SEQUOYAH - UNIT 1 3/4 4-9a Amendment No. 189, 214,
-.. - ~ - ~. ~. ~. - - -.
... ~.. - _.
.. ~... -
. - - --..._-. -._~.-
SURVEILLANCE REQUIREMENTS (Continued) where:
j Vunt upper voltage repair limit l
Vutt lower voltage repair limit
=
mid-cycle upper voltage repair limit based on Vuunt
=
L time into cycle mid-cycle lower voltage repair limit based on v
=
uut Vuunt and time into cycle At length of time since last scheduled inspection
=
during which Vuat and V at were implemented t
cycle length (the time between two scheduled CL
=
steam generator inspections) structural limit voltage Vn.
=
average growth rate per cycle length Gr
=
95-percent cumulative probability allowance for NDE
=
nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC) l 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 methodology in GL 90-05 as supplemented. Vunt may differ at the TSPs and flow distribution baffle.
l l
i i
SEQUOYAH - UNIT 1 3/4 4-9b Amendment No, 189, 214
__. -. -... _.. ~. _.. _ _. _. _ _ -. - _ _... _ _ ~ - _ _... _ _ _ _ _, _ _ _.. _ _.. _ _. _.
REACTOR COOLANT SYSTEM SURVEILLANCE. REQUIREMENTS (Continued) b.-
The' steam generator chall 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 ReDorts 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 returning the steam generators to service should any of the following conditions arise:
1.
If estimated leakage based on the projected end-of-cycle (or if-not practical using the actual measured end-of-cycle) i voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main j
steam line break) for the next operating cycle.
i 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 net practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 X 102, notify the NRC and provide an assessment of the safety significance of the occurrence.
t I'
1 SEQUOYAH - UNIT 1 3/4 4-10 Amendment No. 36, 214,
~
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reaator Coolant System leakage shall be limited te:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, c.
150 gallons per day of primary-to-cecondary 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 i 20 psig.
f.
1 GPM leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coclant 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 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, and leakage from Reactor Coolant System Pressure Isolation Valves, 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 I
hours.
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 R16 each of the above limits by:
SEQUOYAR - UNIT 1 3/4 4-14 Amendment No. 12, 214,
m i
REACTOR COOLANT SYSTEM BA9ES I
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 l
negligible corrosion of the steam generator tubes.
If the secondary coolant a
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-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 i
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 4
GL-95-05 and are applicable only to Westinghouse-designed steam generators (S/Gs) with outside diameter stress corrosion cracking (ODSCC) located at the j
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 4
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 subsegaent 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 voltage repair limit; Vent, is determined from the structural voltage limit by applying the following equation:
Vuat = Vst - Von - Vuog where Vca represents the allswance for flaw growth between inspections and Vuos represents the allowance tor 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.
i SEQUOYAH - UNIT 1 B 3/4 4-3 Amendment No. 36, 189, 214,
/
REACTOR COOLANT SYSTEM BASES l
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 l
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 only where alternate plugging criteria is being applied.
For the purposes of j
this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the 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 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 boundary (tube end up to the start of the tube-to-R193 tubesheet weld). The tube end to tube-to-tubcsheet 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.
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, these results will be promptly reported to the 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 care-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 nr,cessary.
SEQUOYAH - UNIT 1 B 3/4 4-4 Amendment No. 36, 189, 214,
-._.- = -
}
3/4.4.6 REACTOR C0OLANT SYSTEM LEAKAGE I
i 3 /4. 4. 6.1 LEAKAGE DETECTION SYOTEMS The RCS leakage detection systems required by this specification are i
1 provided to monitor and detect leakage from the Reactor Coolant Pressure J
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 fortion of this leakage can be reduced to a threshold value of less than 1 GPM.
This threshold value it sufficiently low to ensure early detection of additional leakage.
i Tha surveillance requirements for RCS Preseure Isolation Valves provide added assarances of valve integrity thereby reducing the probt.oility 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 limit.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known se>urces whose presence will not interfere with the detection of UNIDENTIFIED LEAY. AGE by the leakage detection systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reaccor coolant pump seals exceeds 40 GPM with the modulat;ng l
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 wf'.1 minimize the potential for a significant leakage event during steam line braak.
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 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 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.
SEQUOYAH - UNIT 1 B 3/4 4-4a Amendmet No. 36, 189, 214,
- ~
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
A i
1.
All nonp' lagged tubes that previously had detectable wall pene-trations (greacer than 20%).
1 2.
Tubes in those areas where experience has indicated potential problems 3.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.
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 i
outages.
c.
The tubes selected as the second and third samples (if required by rable 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions or the tubes where imperfections were previously found.
Note:
Tube degradation identified in the portion of the tube that R181 is not a reactor coolant pressure boundary (tube end up to the start of the tube-to-tubesheet weld) is excluded from the Result and Action Required in Table 4.4-2.
d.
Implementation of the steam generator tube / tube support plate repair j
criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 1
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 cat.egories:
Cateoorv Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
SEQUOYAH - Unit 2 3/4 4-11 Amendment No. 181, 211
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this Specification:
1.
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 con-sidered as imperfections.
2.
Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
3.
Degradqp Tuh? vaans a tube contaitiny imperfections greater than or equal to 20% of the nominal tall f.h ekness caused by degradation.
4.
% Degradation means the percentage vf the tube wall thickness affected or removed by degradation.
5.
ppfect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.
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 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 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.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
i 9.
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 prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10.
Tube SuoDort Plate Pluacina Limit 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 SEQUOYAH - UNIT 2 3/4 4-13 Amendment No. 181, 211,
REACTOR COOLANT SYSIEM SURVEILLANCE REQUIREMEhTS (Continued) i a
tube support plates. At tube support plate intersections, the i
. plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:
j 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 to
{
outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1), will i
l be repaired or plugged, except as noted in 4.4.5.4.a.10.c below.
c.
Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion-cracking within the bounds of the tube 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 does not detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion-cracking degradation with 3-a bobbin coil voltage greater than the upper voltage repair j
limit (Note 2) will be plugged or repaired.
d.
Not applicable to SQN.
1 I
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.
i The mid-cycle repair limits are determined from the following equations:
V,e V".
I CL~^ U)
- 1. 0 + NDE + Gr CL i
Vmg* V
~ (V
=Vggg) m yaz SEQUOYAH - UNIT 2 3/4 4-14 Amendment No. 28, 211,
_ _._ _.._ _~
1 l
1
-REACTOR COOLANT SYSTEM 4
j SURVEILLANCE REQUIREMENTS - (Continued) where:
l V n..
u upper voltage repair limit
=
Vutt lower voltage repair limit
=
1 i
mid-cycle upper voltage repair limit based on time I
Vaunt 4
=
.into cycle i
mid-cycle 1cwer voltage repair limit based on v and v tat
=
u uuat i
time into cycle i-g At length of time since last scheduled inspection during which Vuat and Vutt were implemented CL cycle length (the time between two scheduled steam
=
generator inspections) 1 i
V structural limit voltage
=
3t j
Gr average growth rate per cycle length
=
2-s NDE 95-percent cumulative probability allowance for
=
nondestructive examination uncertainty (i.e., a value
)
of 20-percent has been approved by NRC) 4 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 j
tubing or 2.0 volts for 7/8-inch diameter tubing.
I Note 2:
The upper voltage repair limit is calculated according to the methodology in GL 90-05 as supplemented. V may differ at the TSPs ogt j
and flow distribution baffle.
I i
4 b.
The steam' generator shall be determined OPERABLE after completing the I
j corresponding actions (plug all tubes exceeding the plugging limit
)
and all tubes containing through-wall cracks) required by j
j Table 4.4-2.
1 4.4.5.5 Reports l
a.
Following each inservice inspection of steam generator tubes, the j
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 i
shall be submitted to the Commission in a Special Report pursuant to l
Specifiestion 6.9.2 within'12 months following the completion of the inspection.
This Special Report shall include:
2 1.
Fumber and extent of tubes inspected.
SEQUOYAH - UNIT 2 3/4 4-14a Amendment No. 28, 211,
<l
- I C
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
i 3.
Identification of tubes plugged.
}
i c.
Results of steam generator tube inspections which fall into Category R28 l
C-3 shall be reported pursuant to Specification 6.6.1 prior to resumption of plant operation. The written followup of this report i
shall provide a description of investigations. conducted to determine cause of the tube degradation and corrective measures taken to l
prevent recurrence.
i d.
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returning the steam generators to service should any of the following conditions i
arise:
If estimated leakage based on the projected end-of-cycle (or if i.
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 l
tube support plate intersections.
3.
If indications are ident3fied that extend beyond the confines of the tube support plate.
l 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 102, notify the NRC and provide an assessment of the safety significance of the occurrence.
i i
SEQUOYAH -. UNIT 2 3/4 4-14b Amendment No. 28, 211,
l REACTOR COOLANT SYSTEM OPERATIONAL 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 UNIDEN'I'IFIED 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.
l 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 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 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, and leakage from Reactor Coolant System Pressure Isolation Valves, 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 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 each of the above limits by:
SEQUOYAH - UNIT 2 3/4 4-18 Amendment No. 211,
.,I*
REACTOR COOLAN'I' SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be 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 ineervice 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 (prima ry-to-secondary leakage = 150 gallons per day per steam generator),
Cracks having l
primary-to-secondary leakage less than this limit during operation will have d
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 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.
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 for tubing material properties at 6100F (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 is determined from the structural voltage limit by voltage repair limit; VURL, applying the following equation:
Vual " Vst - Vca - Vuos SEQUOYAH - UNIT 2 B 3/4 4-3 Amendment No. 181, 211,
)
\\
REACTOR COOLANT SYSTEM BASES where v represents the allowance for flaw growth between inspections and V a
um 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 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 only where alternate plugging criteria is being applied.
For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the 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 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 boundary (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.
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, these 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, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
SEQUOYAH - UNIT 2 B 3/4 4-3a Amendment No. 181, 211,
- ~,. ~ _ - -
., e:
REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS Tha 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 j
Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection 1
Syr,tems," 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 thereby 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 limit.
l 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.
l 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 i
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 l
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 from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 3.7 gpm.
l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
i Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
I i
SEQUOYAH - UNIT 2 B 3/4 4-4 Amendment No. 211,