ML20216F181
| ML20216F181 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/04/1997 |
| From: | Kennedy J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216F187 | List: |
| References | |
| NUDOCS 9709110180 | |
| Download: ML20216F181 (25) | |
Text
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UNITED STATES s"
j NUCLEAR REGUL ATORY COMMISSION 2
WASHINGTON, D.C. 30686-4e01
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HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET N0. 50-498 SOUTH TEXAS PROJECT. UNIT 1 AMENDMENT TO FACILITY OPERATIMG LICENSE Amendment No. 90 License No. NPF-76 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Houston Lighting & Power Company *
(HL&P) acting on behalf of itself and for the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL),
and City of Austin, Texas (C0A) (the licensees), dated May 17, 1996, as supplemented June 14,1996, March 17, July 29, and July 30, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Ccmmission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- Houston Lighting & Pcwer Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texas and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
9709110180 970904 PDR ADOCK 05000498 P
1
, 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2)llows:of Facility Operating License No. NPF-76 is hereby amended to read as fo 2.
-Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 90, and the Environmental Protectic_.. Plan contained in-Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION dAud b, b r Janet L. Kennedy, Project Manager Project Directorate IV-1 Division of Reactor Projects Ill/IV Office of Nu;1 ear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date-of Issuance:
September 4, 1997
[p?U0 0
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UNITED STATES s
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 3066 Hoot
... +
HOUSTON L1_GHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-499 SOUTH TEXAS PROJECT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.77 License No. NPF-80 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Houston Lighting & Power Company *
(HL&P) acting on behalf of itself and for the City Public Service Board of San Antonio (CPS),-Central Power and Light Company (CPL),
and City of Austin, Texas (C0A) (the licenseos), dated May 17, 1996, as supplemented June 14, 1996, March 17, July 29, and July 30, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- Houston Lighting & Power Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texas and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment 4
and Paragraph 2.C.(2) of Facility Operating License No. NPF-80 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised 4
a through Amendment No. 77, and the Environmental Protection Plan j
contained in Appendix B, are hereby incorporated in the license.
i The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
{
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 7
QAd b.
Janet L. Kennedy, Project Manager l
Project Directorate IV-1 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 4, 1997 i
i t
4 ATTACHMENT TP llCENSE AMENDMENT N05.90 AND 77 TACILITY OPERATING llCENSE NOS. NPF-76 AND NPF-80 DOCKEf NOS. 50-498 AND 50-493 Replace the following sages of the Appendix A Technical Specifications with the attached pages.
T1e revised pages are identified by Amendment number and contain marginal lines indicating tlie areas of change.
The corresponding overleaf pages Are also provided to maintain document completeness.
R@QE INSERT 3/4 4-12 3/4 4-12 3/4 4-13 3/4 4-13 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-16a 3/4 4-16a 3/4 4-16b 3/4 4-16b 3/4 4-18 3/4 4-18 3/4 4-18a 3/4 4-20 3/4 4-20 8 3/4 4-2a B 3/4 4-2a B 3/4 4-3 8 3/4 4-3 B 3/4 4-3a B 3/4 4-3a B 3/4 4-4 B 3/4 4-4 B 3/4 4-5 8 3/4 4-5
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REACTOR COOLANT SYSTEM y titr VAtvrs SURVElltANCE Rt0VIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each NRV shall be demonstrated OPERABLE at least once per 18 months by:
a.
Performing a CHANNEL CALIBRATION on the actuation channel, and b.
Operating the valve through one complete cycle of full travel.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b. or c.
I in Specification 3.4.4.
I s
e I
'l 50lJTH TEXAS - UNITS 11' t 3/44-11 Unit 1 - Amendment No. 44 55 Unit 2 - Amendment No.
s 007 7 1993 m
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REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR LIMITING CONDITION FOR 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 generator (s) to OPERABLE status prior to increasing T,, above 200'F.
SURVEILLANCE RE0VIREMENTS 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 and 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 generator tube minimum sample size, inspection result classification, and the corresponriing action required shall be as specified in Table 4.4-2 and Table 4.4-3.
The inservice inspection of steam generator tubes st.all be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
When applying the exceptions of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection. The tubes selected for each inservice inspection shall include at least 3% of the total number of nonrepaired tubes in all steam generators and 20% of the total number of repaired tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall Se from these critical areas; b.
The first sample of tubes selected for each inservice inspection (subsequent to the oraservice inspection) of each steam generator shall include:
1)
All nonplugged tubes that previously had-detectable wall penetrations (greater than 20%),
2)
Tubes in those areas where experience has indicated potential problems, and SOUTH TEXAS - UNITS 1 & 2 3/4 4-12 Unit 1 - Amendment No.90 Unit 2 - Amendment No.77
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 3)
A tube inspection (>ursuant to Specification 4.4.5.4a.8) shall be performed on eaci selected tube.
If any selected tube doet not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
4)
Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling
- outages, c.
The tubes selected as the second and third samples (if required by Table 4.4-2 or Table 4.4-3) 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 11 (of Technical Specification 4.4.5.4a) shall be 1
inspected via the rotating pancake coil (RPC)from eddy current addy current method over the F* distance. Such tubes are exempt inspection over the portion of the tube below the F* 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 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 fasults of each sample inspection shall be classified into one of the following three categories.
Cateoory Insoection Results C-1 Less than 5% of the total-tubes inspected are degraded tubes and none of the inspected tubes are defective.
SOUTH TEXAS UNITS 1 & 2 3/4 4-13 Unit 1 - Amendment No. Sh&,90 Unit 2 - Amendment-No. 77
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE RE0VIREMEN1$ (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 Nore than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-13a Unit 1 - Amendment No. 83
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVElllANCE RE001REMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this specification:
1)
Tubino or Tube means that portion of the tube or sleeve which forms the primary system to secondary system pressure boundary; 2)
Imoerfection meant ta e n tion to the dimensions, finish, or contour of a tube fm 9d required by fabrication drawings or l s)ecifications.
Eddy-ru,+ent testing indications below 20% of tie nominal tube wall thickness, if detectable, may be considered as imperfections; 3)
Dearadationmeansaservice-inducedcracking,
wastage, wear,orl general corrosion occurring on either inside or outside of a tube; 4)
Dg.gnded Tube means a tube containing imperfections greater l
than or equal to 20% of the nominal wall thickness caused by degraO.d on; 5)
% Dearadation means the percentage of the tube wall thickness l
affected or removed by degradation; 6)
Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube containing a defect is defective; 7)
Pluaaina limit or Reoair limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection, The plugging or repair limit imperfection depths are specified in percentage of the nominal wall thickness as follows:
a.
original tube wall 40%
b.
Westinghouse laser welded sleeve wall 40%
For Unit 1, this definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.12 for the repair litit applicable to these intersections.
8)
Unserviceable describes the condition of a tube if it leaks or l 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; 9)
TubeInspectionmeansaninspectionofthesteamgeneratortubel from the )oint of entry (hot leg side) completely around the U-bend to tie top support of the cold leg; and SOUTH TEXAS - UNITS 1 & 2 3/4 4-15 Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No. 77 l
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE RE0VIREMENTS (Continued)
- 10) Preservice Insoection means an inspection of the full length of l 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 )rior to initial POWER OPERATION using the equipment and teciniques expected to be used during subsequent inservice inspections.
11)
F* criteria fFor Unit 1 on1v1 Tube degradation below a l
s)ecified distance from the hard roll contact point at or near tie top-of-tubesheet (the F* distance can be excluded from considerationtotheacceptancecriter)iastatedinthissection (i.e., plugging of such tubes is not required). The methodology for determination for the F* distance as well as the list of tubes to which the F* criteria is not applicable is described in detail in Topical Report - BAW 10203P, Revision O.
12)
For Unit 1, Tube Sucoort Plate Pluaaina limit is used for the l
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 (repair) limit is based on maintaining steam generator tube serviceability as described r
below:
a)
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage 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.12.c l
- 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 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 outside diameter stress corrosion cracking degradation with bobbin voltage greater than the upper voltage repair limit (Note 2) will be plugged or repaired.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-16 Unit 1 - Amendment No. Ser43,90 Unit 2 - Amendment No. 77
SURVEILLANCE RE0VIREMENTS (Continued) i d)
Certain intersections as identified in Framatome i
Technologies, Inc. Topical Report BAW-10204P, " South Texas i
Project Tube Repair Criteria for ODSCC At Tube Support Plates' will be excluded from application of the I
voltage-based repair criteria as it is determined that i
these intersections may collapse ce deform following a r
e)
If an unscheduled mid-cycle inspection is serformed, the mid-cycle repair limits apply instead of t1e limits j
identified in 4.4.5.4.a.12.a. 4.4.5.4.a.12.b. and 4.4.5.4.a.12.c.
The mid-cycle repair limits will be determined from the equations for mid-cycle repair limits i
of NRC Generic Letter 95-05, Attachment 2, page 3 of 7.
1 Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.12.a.
4 4.4.5.4.a.12.b and 4.4.5.4.a.12.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 methodologyinGenericLetter95-8)assu)plemented.is calculated according to j
V may m
j differ at the TSPs and flow distribution aaffle.
- 13) Tube Reonir refers to a 3rocess that reestablishes tube serviceability. Accepta)1e tube repair will be performed in accordance with the methods described in Westinghouse Reports WCAP-13698, Revision 2, " Laser Welded Sleeves for 3/4 Inch Diameter Tube feedring-Type and Westinghouse Preheater Steam Generators," April 1995 and WCAP-14653, " Specific Application of Laser Welded Sleeves for South Texas Project Power Plant 2
l Steam Generators," June 1996, including post-weld stress relief; 4
Tube repair. includes the removal of plugs that were previously l
installed as a r.orrective or preventive measure. A tube i
inspection per 4.4.5.4.a.9 is required prior to returning previously plugged tubes to service, b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) required by Table 4.4-2 and Table 4.4-3.
4.4.5.5 Reports a.
Within 15 days following the completion of each inservice inspection i
ofsteamgeneratortubes,thenumberoftubespluggedorrepairedinl each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; SOUTH TEXAS - UNITS 1 & 2 3/4 4-16a Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No. 77
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE RE0VIREMENTS (Continued) b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.
This Special Report shall include:
1)
Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged or repaired.
l c.
Results of steam generator tube inspections which f all into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption 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 l 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 identified that extend beyond the confines l
of the tube support plate.
4)
If indications are identified at the tube support plate l
elevations that are attributable to primary water stress corrosion cracking.
5)
If the calculated conditional burst probability based on the l
projected end-of-cycle (or if not practical, using the actu measured end-of-cycle) voltage distribution exceeds 1 x 10',a1 notify the NRC and provide an assessment of the safety significance of the occurrence.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-16b Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No. 77
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENEPATORS TO BE INSPECTED DURING INSEkVICE INSPECTION g
Preservice Inspection No Yes No. of Steam Generators per Unit Two Three Four Two Three Four First Inservice Inspection All One Two Two Second & Subsequent Inservice Inspections One' One' One One' z
TABLE NOTATIONS 1.
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N t of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such ciretanstances the sample sequence shall be modified to inspect the most severe conditions.
2.
The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in 1 above.
3.
Eacht of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and mut u gwat inspections shall follow the instructions described in 1 above.
SOUTH TEXAS - UNITS 1&2 3/4 4-17
Tatde 4.4-2 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Sire Result Acton Required Result Acton Reguwed Result Acton Resumed A nunirnum of
.C-1 None N.A.
N.A.
N.A.
N.A.
S Tubes por C-2, Plug or repair C-1 None N.A.
N.A.
g-S.G.
defecewe tubes and C-2 Plug or reper defective C-1 None
.g inopoet adsstional 2S mbes and inspect tubes in this S.G*
C-2 Plug or repair defective g
adsmonal 45 tubes in mbes this S.G.
C-3 Perform action ser C-3 result of first semple C-3 Perform accon for C-3 N.A.
N.A.
result of first semple C-3 Inspect all tubes in As other th.s S.G., plug or -
S.G.s are None N.A.
N.A.
l.
reper defecove C-1 tubes and inspect 2S tubes in each other Some Perform accon for C-2 N.A.
N.A.
S.G.
S.G.s C-2 result of second sample but no Noefication to NRC addisonei pursuant to 50.72 S.G. are Itd(2) of 10 CFR Part C-3 50 Additional Inspect all tubes in each S.G. is C-3 S.G. and plug or repeer
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defectvve tubes.
Nonficaton to NRC N.A. ~
N.A.
pursuant to 50.72 Itd(21 of 10 CFR Part 50 e
where N es the number of steam generators m the unet, and n es the number of steam generators = _, _ M b -Q S=3 #%
an inspecton.
n SOUTH TEXAS - UNITS 1 & 2 3/4 4-18 Unit 1 - W No.%
Unit 2 - Amendment No.77
Tatde 4.4-3 STEAM GDERATOR FEPAIRED TUGE INSPECTION 1ST SAMPLE INSPECliON 2ND SAMPLE INSPECTION Sample Size Roeult Acten Required Resent Aceien Required A minwnum of 20%
C-1 None N.A.
N.A.
of repened tubes
- C-2 Plus defecove reposed C-1 None tubes and inspect 100% of C-2 Plus defeceve reposed the repened tuties in e S.G.
C-3 Perfenn acean for C-3 result of first semple C-3 inspect as repaired tubes in AN other S.G.s are 90ene this S.G., plus defective C-1 repened tubes and inspect 20% of the repened tubes Some S.G.s C-2 Perform acton for C-2 in each other S.G.
tsut no additionsi result of first semple S.G. are C-3 Notdicamon to NRC pursuant to 50.72 lbH2) of Addisonal S.G. is Inspect as repaired tubes in 10 CFR Part 50 C-3 each S.G. and plus desaceve repened tubes.
Notfication to NRC pursuant to 50.72 RsN2) of 10 CFR Part 50
- Each repair method is' considered a separate populeton for deternunecon of scope esponsion.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-18a Unit 1 - Amendment No. 90 Unit 2 - Amendment No. 77 i
L_.
4 REACTOR COOLANT SYSimM
^
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 Instrumentation shall be OPERABLE:
a.
One Containment Atmosphere Radioactivity Monitor (gaseous or particulate),and b.
The Containment Normal Sump Level and Flow Monitoring System.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a.
With the required containment atmosphere radiosctivity monitor inoperable perform the following cetions or be in at least HOT STf,NDBY 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 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s:
1)
Restore one containment atmosphere monitoring system to OPERABLE status within 30 days and, 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 />, or 3)
Perform a Reactor Coolant System water inventory balance at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
With the required containment normal sump level and flow monitoring system inoperable perform the following actions 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 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s:
1)
Restore the containment normal sump and flow monitoring system to OPERABLE status within 30 days and, 2)
Perform a Reactor Coolant System water inventory balance at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, c.
With both a and b. inoperable, enter 3.0.3.
SURVEILLANCE RE0VIREMENTS 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 Unit 1 - Amendment No. 86 Unit 2 - Amendment No. 73
___-------u
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE i
i LIMITING CONDITION FOR OPERATION j
3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, l
b.
1 gpm UNIDENTIFIED LEAKAGE, i
c.
150 gallons per day of primary-to-secondary leakage through any one j
steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.
0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table i
3.4-1.*
i APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
l 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 />.
I-1 b.
With any Reactor Coolant System leakage greater than any one of the l
above limit., excluding PRESSURE BOUNDARY LEAKAGE and leakage from t
Reactor Coolant System Pressure Isolation Valves, reduce the leakage l
rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY 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 le..t 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 />.
- 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 pressure differential to the one-half power.
Unit 1 - Amendment No. 64,90 SOUTH TEXAS - UNITS 1 & 2 3/4 4-20
- Unit 2 - Amendment No. 77
l REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued)
Manual control of the block valve to:
C.
allow it to be used for manual control o(1) unblock an isolated PORV to f reactor coolant system pressure (Item A), and (2) isolate the PORY with excessive seat leakage (! tem 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 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 modificat<on of Regulatory Guide 1.83 Revision 1.
Inservice inspectionofsteamgeneratortubingisessentialinordertomaintain surveillance of the conditions of t,1e tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions tilat 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 minimize 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 3.4.6.2.c limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System.
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads im)osed during normal operation and by postulated accidents. Operating plants save 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 shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.
Defective tubes may b6 repaired by a Westinghouse laser welded sleeve. The technical bases for sleeving repair are described in Westinghouse Reports WCAP-13698, Revision 2
" Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators," April 1995 and WCAP-14653, " Specific Application of Laser Welded Sleeves for South Texas Project Power Plant Steam Generators," June 1996.
Wastage-type defects are unlikely with pro >er chemistry treatment of the secondary coolant. However, even if a defect s1culd develo) in service, it will be found during scheduled inservice steam generator tu>e examinations.
Except as discussed below, plugging or repair will be required for all tubes with imperfections exceeding the plugging or repair limit of 40% of the original tube nominal wall thickness.
If a tube contains a Westinghouse laser welded sleeve with imperfection exceeding 40% of nominal wall thickness, it must be plugged. The basis for the sleeve plugging limit is based on Regulatory Guide 1.121 analysis, and is described in the Westinghouse sleeving technical reports listed above. Steam generator tube inspections of operating SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2a Unit 1 - Amendment No. 55,02,03,90 Unit 2 - Amendment No. 44,77
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) 31 ants have demonstrated the capability to reliably detect degradation that 1as penetrated 20% of the original tube wall thickness. Repaired tubes are also included in the inservice tube inspection program.
Exclusion of certain areas of Unit I 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 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.
The remainder of the tube belowtheF* distance,isconsiderednotstructurallyrelevantandisexcluded from consideration to the customary plugging criteria of 40% throughwall.
The amount of primary to secondary leakage from tubes left in service by a) plication of the F* criterion has been determined by verification testing.
T.11s 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 connc;uences as a result of this leakage is less than the 10 CFR 100 limits.
I i
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 (ODSCC) 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 apply only to indications where the i
degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the thickness of the support plate.
Refer to GL 95-05 for i
additional description of the degradation morphology, Implementation of SR 4.4.5 requires a derivation of the voltage i
structural limit from the burst versus voltage empirical correlation and then l
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.
air limit; V is determined from the structural The upper voltage re$ying the fol$w,ing equation:
voltage limit by app l
Y
-Vn-V,-V where V represent the allowance for flaw growth between inspections and V represenks the allowance for potential sources of error in the measurementN c
the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05, d
SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3 Unit 1 - Amendment No. Sh43,90 Unit 2 - Amendment No. 77
REACTOR COOL ANT SYSTEM BASES STEAM GENERATOR $ (Continued)
The mid-cycle equation in SR 4.4.5.4.a.12.e should only to. used during l
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 practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service. Note that if leakage and conditional burst probability were calculated using the EOC 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 EOC 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
)rovided to monitor and detect leakage from the reactor coolant pressure soundary. 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 unacce) table since it may be indicative of an impending gross failure of the pressure )oundary. 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 )ortion of this leakage can be reduced to a threshold value of less than 1 gpm. T11s threshold vale is sufficiently low to ensure early detection of additional leakage.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3a Unit 1 - Amendment No. 63,90 Unit 2 - Amendment No.77
BASES OPERATIONALLEAKAGE(Continued)
For Unit 1, the leakage limits incorporated into SR 4.4.6 are more restrictive than the standard operating leakage limits and are intended to provide an additional margin to accomodate a crack which might firow at a greater than expected rate or unexpectedly extend outside the th'ekness of the tube support plate. Hence, 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 be detected, and the plant shut down in a timely manner.
For Units 1 and 2, the steam generator tube leakage limit of 150 gpd for each steam generator not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline valves in-the event of either a steam generator tube rupture or steam line break. The-150 gpd limit per steam generator is conservative compared to the assumptions used in the analysis of these accidents.
The 150 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
The 10 g>m IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leacage 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 sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent 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 tiny, verification of valve integrity is required.
Since these valves are impor'; ant in preventing overpressurization.and rupture of the ECCS low pressure pi>ing 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-limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-4 Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No.77
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,how that operation may be continued with containment concentration and fluoride limits are time and temperature dependent.
Corrosion studies s levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.
The Surveillance Requirements provide adequate assurance that 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 BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 150 gpd per steam generator. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site l locations. These values are conservative in that specific site parameters of the STPEGS site, such as SITE BOUNDARY location and meteorological conditions, were not considered inthis 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 l-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 radiciodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT l-131 and because, if the limit is exceeded, theradiciodinelevelistobedetermlnedevery4 hours.
If the gross
-s>ecific activity level and radiciodine level in the reacto" coolant were at tiair limits, the radiciodine contribution would be approximately 1%.
In a release of reactor coolant with a typical mixture of radioactivity, the actual radiciodine contribution would >robably be about 20%. The exclusion of radio-nuclides with half-lives less tian 15 minutes from these determinations has SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-5 Unit 1 - Amendment No.90 Unit 2 - Amendment No.77
_J
t REACTOR COOLANT SYSTEM RA$ES SPECIFICACTIVITY(Continued) l been made for several reasons. The first consideration is the difficulty to identify short lived radionuclides in a sample that requires a significant i
time to collect, transport, and analyze.
The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes decay time.
The choice of 15 minutes for the half-life cutoff was made because of the nuclear
.i characteristics of the typical reactor coolant radioact: ity.
The radionuclides in the typical reactor coolant have half lives of less than 4 minutes or half-lives of greater than 14 minutes which allows a distinction between the radionuclidesaboveandbelowahalf-lifeof15 minutes.
For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perfom the analysis of about 90 minutes.
After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties.
The counter should be reset to a reproducible efficiency versus energy.
It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than I hour, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about I day, about I week, and about 1 month.
4 Reducing T,yg to less than 500*F prevents the release of activity should-a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
A reduction in frequency of isotopic analyses following ;
power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE t1MITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:
1.
The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance.-
with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
SOUTH TEXAS - UNITS 1 & 2 8 3/4 4-6
..