ML20117H163
| ML20117H163 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 08/29/1996 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20117H095 | List: |
| References | |
| NUDOCS 9609090140 | |
| Download: ML20117H163 (20) | |
Text
._
ATTACHMENT 3 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS 3/4.4.5 AND 3.4.6.2 Paaes Attached 3/4 4-12 3/4 4-13
- 3/4 4-13a
- 3/4 4-14
- 3/4 4-15 3/4 4-16 3/4 4-16a 3/4 4-16b 3/4 4-17
- 3/4 4-18 3/4 4-19
- 3/4 4-20 3/4 4-21
- B 3/4 4-2a B 3/4 4-3 8 3/4 4-3a
- B 3/4 4-4 B 3/4 4-5 No changes, provided for completeness NOTE: All revisions are shown by vertical change bar in right-hand column.
Additions are Redlined.
Deletions are shown by strike-out bars I
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l REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS l
LIMITING CONDITIONS FOPS OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
With one or more steam generators inoperable restore the inoperable j
generator (s) to OPERABLE status prior to increasing T above 200 F.
y SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
4.4.5.1 Steam Generator Samole Selection and Insoection - Each steam generator shall be determined OPERABLE during shutdown by selecting End inspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.5.2 Steam Generator Tube Samole Selection and Insoection - The steam
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generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable er the acceptance criteria of Specification 4.4.5.4.
WhifiDiijijiljfhgitheife~~ JdFisMN4f5725tfiFoughW4?572WpfiVfuuildifdcts i
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etiony[The tubes se!d%31eepingfaregioticonsjderydi4DiMen16c inclUd5 aflisst 3%~of~the total number of tubes in all steam generators: the i
tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar )lants with similar water chemistry indicates critical areas to )e inspected, then at least 50% of the tubes inspected shall be from these critical areas:
i b.
The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1)
All nonplugged tubes that previously had detectable wall i
penetrations (greater than 20%),
4 2)
Tubes in those areas where experience has indicated potential problems, and SOUTH TEXAS PROJECT - UNITS 1 & 2 3/4 4-12
J REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 3)
A tube inspection ()ursuant to S)ecification 4.4 5.4a.8) shall be performed on eac1 selected tu)e.
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 ) late voltage-based repair criteria shall be inspected by Jobbin 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. and 2)
The inspections include those portions of the tubes where imperfections were previously found.
d.
For Unit 1. any tube allowed to remain in service per Acceptance Criterion 10 (of Technical Specification 4.4.5.4) shall be ins)ected via the rotating pancake coil (RPC) eddy current method over tie F*
distance.
Such tubes are exemat from eddy current inspection over the portion of the tube below the :* distance which is not structurally relevant.
e.
For Unit 1.
implementation of the steam generator tube / tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot leg and cold leg tube support alate intersections down to the lowest cold-leg tube support plate witi 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:
Cateaory Insoection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
1 SOUTH TEXAS - UNITS 1 & 2 3/4 4-13 Unit 1 - Ainendment No. 82-83
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l REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10%
of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are l
degraded tubes or more than 1% of the inspected tubes are defective.
l l
L Note:
In all inspections, previously degraded tubes must exhibit l
significant (greater than 10%) further wall penetrations to be I
included in the above percentage calculations l
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i SOUTH TEXAS - UNITS 1 & 2 3/4 4-13a Unit 1 - Amendment No. 83
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Insoection Freauencies - The above required inservice inspections of
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steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of l
initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 1
calendar months after the previous inspection.
If two consecutive inspections, not including the preservice inspection, result in i
all inspection results falling into the C-1 category or if two I
consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals l
fall in Category C-3. the inspection frequency shall be increased i
to at least once per 20 months. The increase in inspection l
frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a: the interval may then be extended to a maximum of once per 40 months; and c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent j
to any of the following conditions:
1)
Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 3)
A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 4)
A main steam line or feedwater line break.
i SOUTH TEXAS - UNITS 1 & 2 3/4 4/14 i
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STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceotance Criteria a.
As used in this specification:
1)
Imoerfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or s )eci fications.
Eddy-current testing indications below 20% of tie nominal tube wall thickness, if detectable, may be considered as imperfections:
2)
Dearadation means a service-induced cracking,
wastage, wear, or general corrosion occurring on either inside or outside of a tube; 3)
Dearaded Tube means a tube containing imaerfections greater than or equal to 20% of the nominal wall thic(ness caused by degradation:
4)
% Dearadation means the percentage of the tube wall thickness affected or removed by degradation:
5)
Defect means an imperfection of such severity that it exceeds the plugging Q eph{h limit. A tube containing a defect is I
defective:
6)
Pluaaina Limit BPRs6sTFTiiiif means the imperfection depth at orbeyondwhichthetubesha'lberemovedfromservicebf plugg1hi Ifsissl T
mayif*:o@jme unsery%bfHi&RT@illhlthsiiffsetsd!IFiaibeciUMig 1ceableipn, ::oit6etnextMnspectiohEThei percentag(ogrepa1@imittihdectio%depthstareispecified@_n, plugg.i.ng elof4ths3nominalj[iva.lgthis:knessiasifo]Jows; and =
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gr;tPS~6fij@i6sWtUbsislU"jj? ""?IMW*"P"*"*T40%
EEajffi ouiegechno]N d$$$p[]M(1,a6,d@%
For Unit 1 this definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being a) plied.
Refer to 4.4.5.4.a.11 for the repair limit applica)le 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 Specification 4.4.5.3c., above:
8)
Tube Insoection means an inspection of the steam generator tube from the Joint of entry (hot leg side) completely around the U-bend to tie top support of the cold leg: and SOUTH TEXAS - UNIT 1 & 2 3/4 4-15 Unit 1 - Amendment No. 83 l
STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) l 9)
Preservice Insoection 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 i
the tubing.
This inspection shall be performed prior to initial 1
POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10)
.F* criteria (For Unit 1 only) Tube degradation below a s)ecified distance from the hard roll contact point at or near t1e top-of-tubesheet (the F* distance) can be excluded from consideration to the acce)tance criteria stated in this section (i.e. plugging of such tu>es is not re For sleeves extending into 'the P: distance.'the^F* quired). distance' for'the sleeved tube will start below the bottom joint of the sleeve / ~The methodology foF* determination for the'F* ' distance'as~well as the list of tubes to which the F* criteria is not a) described in detail in Topical Report - BAW 102)plicable is 3P. Revision 0.
11)
For Unit 1, Tube Suocort Plate Pluaaina 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 tube support plates. At tube support plate intersections, the plugging (re) air) limit is based on maintaining steam generator tu)e serviceability as described below:
a)
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube suoport plate with bobbin voltage less thanoregualtotheiowervoltagere will be allowed to remain in service. pair limit (Note 1),
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 volta greater than the lower voltage repair limit (Note 1)ge, will be repaired or plugged, except as noted in 4.4.5.4.a.11.c below.
c)
Steam generator tubes with indications cf 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 the upper repair voltage 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 bobbin voltage greater than the upper voltage repair limit (Note 2) wi11 be plugged or repaired.
SOUTH TEXAS - UNIT 1 & 2 3/4 4-16 Unit 1 - Amendment No. 83
REACTOR COOLANT SYSTEM STEAM GFNERATORS SURVEILLANCE REQUIREMENTS (Continued) d)
Certain intersections as identified in Framatome Technologies. Inc. Topical Re) ort BAW-10204P. " South Texas Project Tube Repair Criteria or ODSCC At Tube' Support r
Plates" will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA + SSE event.
e)
If an unscheduled mid-cycle inspection is aerformed, the I
mid-cycle repair limits apply instead of tie limits identified in 4.4.5.4.a.11.a. 4.4.5.4.a.11.b. and i
4.4.5.4.a.11.c.
The mid-cycle repair limits will be determined from the equations for mid-cycle repair limits of NRC Generic Letter 95-05. Attachment 2. page 3 of 7.
Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.11.a.
4.4.5.4.a.11.b. and 4.4. 5.4.a.11.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 (V is calculated according to the metnodology in Generic Letter $)5-05 as su)plemented.
i V
m may differ at the TSPs and flow distribution aaffle.
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The steam generator shall be determined OPERABLE after completing the corresponding actions (p ug Krie~pa))ing through-wall cracks) plugging all tubes exceeding the orgjreW h limit and ail ubes contain required b'y^TaYs"4.4-2.
4.4.5.5 Reoorts Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged 6CF#pa" ired in I
i each steam generator shall be reported to the Commissibn in a $pbcial Report pursuant to Specification 6.9.2:
SOUTH TEXAS - UNITS 1 & 2 3/4 4-16a Unit 1 - Amendment No. 83
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REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) i 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 includ?:
1)
Number and extent of tubes inspected.
2) location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged @Ejiijjjid.
I 1
Results of steam generator tube ins)ections which fa?1 into Category c.
C-3 shall be reported in a Special Report to the Comm1:sion pursuant to Specification 6.9.2 within 30 days and prior to restr6ption of plant operation. 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 Unit.1. 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) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the line break) for the next operating cycle. postulated main stream 2.
If circumferential crack-like indications are detected at the tube support plate intersections.
3.
If indications are identified that extend beyond the confiries of the tube support plate.
4.
If indications are identified at the tube support plate i
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'.
notify the NRC and provide an assessment of the safety significance of the occurrence.
1 SOUTH TEXAS - UNITS 1 & 2 3/4 4-16b Unit 1 - Amendment No. 83
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.x TABl.E 4.4-1 Q
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MINtMUM NUMBER OF STEAM GENER ATORS TO BE w
INSPECTED DUntNG INSERVICE INSPECTION n
(
(
Preservice Inspection r
No Yes ta)
No. of Steam Generators pee Unit Two Three Four Two Threr Four
'i First inservice Inspection All One Two Two Second & SwLm ~a imervice inspections One' One One2 l
One3 TADt.E NOTATIONS 1, The Imervice Impection moy be flmited to one steem generator on a rotating schedule g,.e,w;;Ing 3 N % of the tubes (where N is the number of steem generators in the planti If the results of the first or previous inspections Indicate that sti steem generstors are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steem generators mey be found to be rnote severe then those in other steam genersters. Under sich circum-i stances the semple sequence shaft be modified to impect the most severe conditlom.
i
- 2. The other steem generster not Impected during the first inservice inspection shatl be Impected. The thin! and subseov Impections shoukt fotfow the Imtructions descritnd in 1 above.
- 3. Each of the other two steem geneestors not impected during the first Imervice Impections shall be I upected durin second and third impections. The fourth and subsequent impections shall foffow the Imtructions describM in 1 above.
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__ TABLE 4 4-2 h
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N N
N 1ST SAMPt.E INSPECTM YD S%PLE INSPECTION
'N3RD SAMPl.E INSPECTION cth Required Re%
tion Required Re\\s Sample Slee Renutt A
Act*m Requ W A minimum of C-1 N. A.
A.
N. A.
N. A.
S Tubespee C-2 Plughett e tubes C-1 tu N. A.
N. A.
and impact idditionel
.(
Phsg'Tdefect tubes C-1 None/
2S tubes in i Nis S. G.
C-2 and impect itional C-2 PlugWecth*: tubes 4S tubes in th S. G.
p g
C-3 C-3 result of first sampk
[
Perform action he 4
C-3 C-3 result of firit N. A.
N. A.
sampk C-3 Irnpect att LL-i Att other I
this S. G., plejhn S. G.s are None N. A.
N. A.
fective tubes and C-t Impect 25 tubes in l
g,; s, c,,
t och othee S. G.
Perform action for N. A.
N. A.
)
C-2 but no C-2 roult of secor'd additionet semph i
Notificatiostle NRC S. G. are F
pursuant to $50.72 C-3 (bl(2) of 10CFR.
Additional - Impect all tubes in Part 50 S. G. h C-3 each S. G. and plugh defective tubes.
Notification to NRC N. A.
N. 'A.
pursuant to $50.72 (bl(2) of 10 CFR Part 50 M
where n h the number of steem eenerators in the unit. and n h the number of steam venerators impetted 3,3 g during an impection n
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REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
a.
The Containment Atmosphere Gaseous Radioactivity Monitoring System.
l b.
The Containment Normal Sump Level and Flow Monitoring System, and c.
The Containment Atmosphere Particulate Radioactivity Monitoring System.
APPLICABILITY: MODES 1. 2. 3. and 4.
ACTION:
a.
With a. or c. of the above required Leakage Detection Systems iroperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for j
gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required Gaseous or Particulate Radioactive Monitoring System is j
inoperable: otherwise, 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 />.
b.
With b. of the above required Leakage Detection Systems inoyerable, 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 SHJTDOWN within i
the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
{
c.
With a. and c. of the above required Leakage Detection Systems inoperable:
1)
Restore either Monitoring System (a. or c.) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and j
i 2)
Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
and 3)
Perform a Reactor Coolant System water inventory balance at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Otherwise, 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 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
a.
Containment Atmosphere Gaseous and Particulate Monitoring Systems performance of CHANNEL CHECK. CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and b.
Containment Normal Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-19 w
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE B0UNDARY LEAKAGE, b.
1 gpm UNIDENTIFIED LEAKAGE, For Unit 1 150 gallons per day of primary-to-secondary leakage throu c.
steam generator. and for Unit 2. I gp-total rc :tcr to ccend:ry le:gh any one
., ge g y ll ste:: generator and 500 gallen: per d y through any onc :te:
n,~._.,
'd.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a e.
Reactor Coolant System pressure of 2235 i 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*
APPLTCABILITY: MODES 1. 2, 3, and 4.
ACTION:
a.
With any PRESSURE B0UND/RY 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 />.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than c.
the above limit, isolate the high 3ressure portion of the affected system from the low pressure portion within 4 lours 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 />.
i o Test _ pressures less than 2235 psig but greater than 150 psig are allowed. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressu"e differential to the one-half power.
I SOUTH TEXAS - UNITS 1 & 2 3/4 4-20 Unit 1 - Amendment No. 83
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCES REQUIREMENTS (Continued) 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a.
Monitoring the containment atmosphere gaseous radioactivity and particulate i
-radioactivity channels at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; b.
Monitoring the containment normal sump inventory and discharge at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
4 c.
Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and d.
Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:
a.
At least once per 18 months, b.
Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 moriths,
c.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve. and d.
Prior to entering MODE 2 following valve actuation due to automatic or manual action or flow through the valve except for valves XRH0060 A, B, C. and XRH0061 A B, C.
e.
As outlined in the ASME Code,Section XI, paragraph IWV-3427(b).
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-21 Unit 1 - Amendment No. 22 Unit 2 - Amendment No. 12
REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued)
C.
Manual Control of the block valve to:
(1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item A), and (2) isolate the PORV with excessive seat leakage (Item B).
D.
Manual control allows a block valve to isolate a stuck-open-PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that l
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 i
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 3rogressive degradation due to design.
manufacturing errors, or inservice conditions tlat lead to corrosion.
Inservice i
inspection of steam generator tubing also provides a means of characterizin and cause of any tube degradation so that corrective measures can be taken.g the nature i
l The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to minimize corrosion of the steam l
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 3.4.6.2.c limitation of l
steam generator tube leakage between the Reactor Coolant System and the Secondary i
Coolant System.
Cracks having a primary-to-secondary leakage less than this limit l
during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage as low as 150 gallons per day per steam generator can readily be detected.
Leakage in excess of this limit will require plant i
shutdown and 4 un and plugged orgrepscheduled inspection guri.ng whici)ididibyMra_matomelTd m
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Wastage-ty)e defects are unlikely with proper chemistry treatment of the secondary coolant.
iowever, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Except as discussed below.
p ugging or repain will be recuired for all tubes with imperfections exceeding the p
ing or, repait limit of 4C% of the original tube nominal wall thickness.
If a' tube ins"a Framatome Technologies'Electroslesve withTim)erfections' eiceeding"20% of the C
i nominal sleeve wall thickness it must be plugged./The ) asis for the sleeve plug 11ng ~~
limit is based on Regulatory Guide 1.121 analysis, and is described in the Framalome l
Technologies technical, reports.~ Steam generator ~ tube 'inspsctions' of o)erating plants have~demonstratea'the capability to reliably detect degradation that 1as )enetrated 20%
I of the original tub.e wall thickness. @hjfedjtubbslafGTg@lsdedyg@@56svge (OtisHg,spgetigg_1programj i
Exclusion of certain areas of Unit 1 tubes from consideration has been analyzed using an F* criteria. The criteria allows service induced degradation deep within the tubesheet to remain in service. The analysis l
l SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2a Unit 1 - Amendment No. 83 j
Unit 2 - Amendment No. 44 i
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) j methodology determines the length of sound fully rolled expanded tubing required in the uppermost area within the tubesheet to preserve needed structural margins for all service conditions. F6hIsleeV i?that9penetratsith6%*sdisthhbur f#thiiphrdntiftubs FfdfstiabsIf6FIthsssitbbe' MilHbesmeasuredifromithbottostofithdBlee o
s The" PemiliidEFb f ~th~i~fub's".~^ be167tWe* F*~ dis t shes."Ti~E6hsi dEF#d ^n6t"~stFG6tiiFsl ly~~Fil eva nt l
and is excluded from consideration to the customary plugging criteria of 40%
l throughwall.
The amount of primary to secondary leakage from tubes left in service by application of the F* criterion has been determined by verification testing. This leakage has been considered in the calculation of postulated primary to secondary leakage under accident conditions.
Primary to secondary leakage during accident conditions is limited such that the associated radiological consequences as a result of this leakage is less than the 10 CFR 100 limits.
For Unit 1. 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 (SGs) with outside diameter stress corrosion cracking (0DSCC) located at the tube-to-tube support plate intersections. The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG. Additionally, the repair criteria a) ply only to indications where the degradation mechanism is dominantly axial ODSCC wit 1 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 1
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: Va. is determined from the structural voltage limit by applying the following equation:
Va-V3t - Vm - Va where Va represent the allowance for flaw growth between inspections and V represents the allowance for potential sources of error in the measurement of the bobbgin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
I SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3 Unit 1 - Amendment No. 83
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)
The mid-cycle equation in SR 4.4.5.4.a.11.e should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning the SGs to service.
For the purpose 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 3ractical to complete these calculations using the projected E0C voltage distri)utions prior to returning the SGs to service.
Note that if leakage and conditional burst probability were calculated using the E0C voltage distribution for the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected E0C voltage distribution should be provided per the GL section 6.b. (c) criteria.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3. these results will be promptly reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days and 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.
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 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.
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.
l l
l l
f SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3a Unit 1 - Amendment No. 83
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)
For Unit 1. the leakage limits incorporated into SR 4.4.6 are more restrutive than the standard operating leakage limits and are intended to provide an additional margin to accommodate a crack which might grow at a greater than ex)ected rate or unexpectedly extend outside the thickness of the tube support plate.
lence, the reduced leakage limit, when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will l
be detected. and the plant shut down in a timely ma,1ner.
For Unit 2. The teta4 steam generator tube leakage limit of gpm W{gpd for all Eithl steam generator not isolated from the RCS ensures that the dosage con ribution froin~the l
tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline l
values in the event of either a steam generator tube rupture or steam line break. The gpm 1503p'Used in thE"Maljsis of thEse accidents.d limit M?itudiiilgenefatbr: is co l
assumpti6ns
~
The 600^150"Wa~1dakacjs~1isit per steam generator ensures that steam generator tube integrity is~iiisintained in the event of a main steam line rupture or under LOCA conditions.
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 specified allowed leakage from any RCS pressure isolation valve is sut ficiently low to ensure early detection of possible in-series check valve failure.
It it apparent i
that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, 1
verification of valve integrity is requirm. Since these valves are important in preventing overpressurization and rupture of the ECCS low )ressure piping which could result in a LOCA that bypasses containment, these valves s1ould be tested periodically to ensure low probability of gross failure.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
3/4.4.7 CHEMISTRY The limitaticns on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the )otential for Reactor Coolant System leakage or failure due to stress corrosion.
iaintaining SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-4 Unit 1 - Amendment No. 83 1
I
}
REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, l
up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time l
interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.
The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE B0UNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guidelines values following a steam generator tube i
rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 4-9pm 150?g_pdW~iitistilseheFit6d. The values for the limits I on specific activity represent Tiiriils based 0pbh~5 pisFs5stric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the STPEGS site, such as SITE B0UNDARY location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
The sample analysis for determining the gross specific activity and E can exclude the radioicdines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131 and because, if the limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the gross specific activity level and radiciodine level in the reactor coolnt were at their limits, the radiciodine contribution would be approximately 1%.
In a release of reactor coolant with a ty)ical mixture of radioactivity, the actual radiciodine contribution would pro) ably be about 20%. The exclusion of radionuclides with half-lives less than 15 minutes from these determinations has SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-5