ML20206A127
| ML20206A127 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/19/1999 |
| From: | Gramm R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20206A130 | List: |
| References | |
| NUDOCS 9904280067 | |
| Download: ML20206A127 (19) | |
Text
_
etc o
t UNITED STATES f
j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20eeH001 o
9..ca/
STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-498 SOUTH TEXAS PROJECT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.107 License No. NPF-76 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by STP Nuclear Operating Company
- acting on behalf of itself and for Houston Lighting & Power Company (HL&P), the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and City of Austin, Texas (COA) (the licensees), dated September 30, 1998, 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; j
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.
'STP Nuclear Operating Company is authorized to act for Houston Lighting & Power Company (HL&P), 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.
9904290067 990419 PDR ADOCK 05000498 P
FM i
. 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) of Facility Operating License No. NPF-76 is hereby amended to read as follows:
2.
Technical Soecifications The Technical Specification:: contained in Appendix A, as revised through Amendment No. 107, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in ac::ordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance, and shall be implemented following the replacement of Unit 1 Model E steam generators with Model A94 steam generators and prior to Unit 1 operation with the A94 steam generators installed.
FOR THE NUCLEAR REGULATORY COMMISSION k
Ah~
Robert A. Gramm, Chief, Sec* ion 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifiertions i
Date of issuance: April 19, 1999 1
/
pMeru
~
p UNITED STATES g
j NUCLEAR REGULATORY COMMISSION O
t WASHINGTON. D.C. enmas met s.;'...../
STP NUCLEAR OPERATING COMPANY DOCKET NO, 50-499 SOUTH TEXAS PROJECT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. NPF-80 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by STP Nuclear Operating Company
- acting on behalf of itself and for Houston Lighting & Power Company (HL&P), the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and City of Austin, Texas (COA) (the licensees), dated September 30, 1998, 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.
- STP Nuclear Operating Company is authorized to act for Houston Lighting & Power Company (HL&P), 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 and Paragraph 2.C.(2) of Facility Operating License No. NPF-80 is hereby amended to read as follows:
2.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 94, and the Environmental Protection 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 and shall be implemented following the replacement of Unit 1 Model E steam generators with Model A94 steam generators and prior to Unit 1 operation with the A94 steam generators ir. stalled.
FOR THE NUCLEAR REGULATORY COMMISSION A A A-Robert A. Gramm, Chief, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: April 19, 1999
ATTACHMENT TO LICENSE AMENDMENT NOS.107 AND c)4 FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 DOCKET NOS. 50-498 AND 50-499 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3/4 4-11 3/4 4-11*
3/4 4-12 3/4 4-12 3/4 4-13 3/4 4-13 3/4 4-13a 3/4 4-13a*
3/4 4-14 3/4 4-14 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 4-18a 3/4 4-18a B 3/4 4-2a B 3/4 4-2a B 3/4 4-3 B 3/4 4-3 B 3/4 4-3a B 3/4 4-3a B 3/4 4-4 B 3/4 4-4 l
1
)
- Overleaf pages provided to maintain document completeness. No changes contained on these pages.
REACTOR COOLANT SYSTEM REl1EF VALVE 5 k
SURVEILLANCE REOUTREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV 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 6PERABLE at least once per 92 days by cperatinfs closed in order to meet the requirements of ACTION b. or c.
the valve through one complete cycle of full travel unless the block valve I
in Specification 3.4.4.
I
.~.
4 t
SOUTH TEXAS - UNITS 11't 3/4 4-11 Unit 1 - Amendment No. 44 55 Unit 1 - Amendment No.
s 0077 1993
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 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 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 corresponding action required shall be as specified in Table 4.4-2 and Table 4.4-3. 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 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 shallinclude at least 3% of the total number of nonrepaired tubes in all steam generators and (for Model E steam generators only) 20% of the l
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 be from these critical areas; 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 penetrations (greater than 20%),
2)
Tubes in those arcas where experience has indicated potential problems, and SOUTH TEXAS - UNITS 1 & 2 3/4 4-12 Unit 1 - Amendment Nc. 90,107 Unit 2 - Amendment No. M,94
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 3)
A tube inspection (pursuant to Specification 4.4.5.4a.9) shall be performed l
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)
For Model E steam generators only, indications left in service as a result of l
)
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 tubs aset 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 Model E steam generators only, implementation of the steam generator l
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 results of each sample inspection shall be classified into one of the following three categories.
Category Insoection Results 1
C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
4 SOUTH TEXAS - UNITS 1 & 2 3/4 4-13 Unit 1 - Amendment No. Ge;e9;90;96,107 Unit 2 - Amendment No. W;e9, 94
o REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEftlANCE REOUIREMENTS (Continued)
C-2 One or more tubes, but not more than IX of the total tubes inspected are defective, or between 5% and 10%
of the total tubes inspected are degraded tubes.
C-3 More than los 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.
)
)
i SOUTH TEXAS - UNITS 1 1 2 3/44-13a Unit 1 - Amendment No. 83
1
\\
{
STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Insoection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection following steam generator replacement shall be l
performed after 6 Effective Full Power Months, but within 24 calendar months of initial criticality after the steam generator replacement. Subsequent inservice l
inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradatiaon has occurred, the inspection interval may be extended to a maximum of once per 40 months; Note:
Inservice inspection is not required during the steam generator replacement outage, b.
If the results of the inservice ir.spection of a steam generator conducted in accordance with Tab!e 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection 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 to any of the following conditions:
1)
Primary-to-seconday 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.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-14 Amendment No.107 Amendment No. 94
I-REACTOR COOLAfJT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Cnteria a.
As used in this specification:
1)
Tubina or Tube means that portion of the tube or sleeve which forms the primary system to secondary system pressure boundary; 2)
Imoerfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; 3)
Dearadation means a service-induced cracking,
wastage, wear, or general corrosion occurring on either inside or outside of a tube; 4)
Dearaded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; 5)
% Dearadation means tne percentage of the tube wall thickness affected or removed by degradation:
6)
Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube contrdning 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 (for Model E steam generators only) 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 Model E steam generators, this definition does not apply to tube support l
plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.11 for the repair limit applicable to these l
intersections.
8)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above; 9)
Tube insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the coldleg; l
SOUTH TEXAS - UNITS 1 & 2 3/4 4-15 Unit 1 - Amendment No. 09;99;96,107 Unit 2 - Amendment No. 97;69,94
p, REACTOR COOLANT SYSTEM STEAM GENERATORS SURVElLLANCE REQUIREMENTS (Continued)
- 10) - Preservice Insoection means an inspection of the fulllength 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.
11)
For Model E steam generators only, Tube Suooort Plate Plucoino Limit is l
used for the disposition of a mill annealed 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 below:
3 1
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).
diameter stress corrosion cracking within the bounds of the tube Steam generator tubes, whose degradation is attributed to outside 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.11.c below.
l 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 j
lower voltage repair limit (Note 1) but less than or equal to the upper j
repak voltage limit (Note 2), may remain in service if a rotating pancake coilihspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking j
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. OE;09;99;96,107 Unit 2 - Amendment No. 77;09, 94
)
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) d)
If an unscheduled mid-cycle inspection is performed, the mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.11.a.
4.4.5.4.e.11.b, and 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.
l Note 1:
The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing.
Note 2:
The upper voltage repair limit (Vunt) is calculated according to the methodology in Generic Letter 95-05 as supplemented. Vunt may differ at the TSPs and flow distribution baffle.
12)
Tube Reoair refers to a process that reestablishes tube serviceability for Model E steam generators only. Acceptable 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 Steam Generators," June 1996, including post-weld stress relief; Tube repair includes the removal of plugs that were previously installed as a corrective or preventive measure. A tube 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 (for Model E steam generators only) 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 of steam generator tubes, the number of tubes plugged or repaired in 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. 09;90;96,107 Unit 2 - Amendment No. M;es,94 I
i
- STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) b.
The complete results of the steam generator tube inservice inspection s.all be j
submitted to the Commission in a Special Report pursuant to Specificati >n G.9.2 1
within 12 months following the completion of the inspection. This Speci l Report shallinclude:
1)
Number and extent of tubes inspected, 2)
Location and percent of wall-thickness penetration for each indic ation of an imperfection, and 3)
Identification of tubes plugged or repaired.
c.
Results of steam generator tube inspections which fallinto Category 0-3 shall be reported in a Special Report to the Commission pursuant to Specific stion 6.9.2 within 30 days and prior to resumption of plant operation. This repe t shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence d.
For Model E steam generators, implementation of the voltage-br sed repair criteria l
to tube support plate intersections, notify the Staff prior to returr eng the steam generators to service should any of the following conditions ari,e:
1)
If estimated leakage based on the projected end-of-cy' <e (or if not practical, using the actual measured end-of-cycle) voltage distrbution exceeds the leak limit (determined from the licensing basis dose calet ution 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 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 104, 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. OS;90;96,107 Unit 2 - Amendment No. 77;09,94
l 704 19
~,,
9 9 M.
.oo NN t tnn ee mm n
f dd o
nn i
ee s
)
e s
2 mm 3
2 b
e h
(
d AA u
b
)
Cle Cle t
u b
e r
p p
d gt C(
eu p
om om i pen 70 r
eudR 2 1 2 r
N q
e e
l f
e f
r t t e
r a
a r
ii O
R A.
n n
o e
ns o
ns a
nn pd ao0 5 i
I 5
UU T
n N
N v
ot ot enpt N
it s
E io i
it s t
r ae nor C
t r
r c
ci ci
.r ot a l
o af af t
l P
n a G. e i P
c t t f
f M
mo mo N
S A
vanR o
tcSi t ica F is O
N s
ol ol e
r r
t t
cf u g
f u f u ph eisC n
I uh I
T a
cf t r r s r
C E
s s
e n ae ou0 p
e e e
Mt P r P r i e d N p1 x
L E
P e
P S
M e
N A
p S
e o
I r
s c
E a
2 la f
i s
B D
- n U
N s.
Co G.
o t
G.
it 3 S
n T
2 l
u
- s. d 1
2 3
A.
i -
S o
s D
e N
C C
C G. d C la i
r t
E R
e Sae n
a R
h oa it r
o n
I i
t e n.
d 3 r
A o
mt G.
d -
e a
i m
3-P l
1 4
E l
ou t
8 R
1 4
d R
4 e
r O
o 4
l f
n f
b T
o i
t s o
/
f a
s s
ce n
3 T
A
%i eeeb 2
it
)
o R
h E
d d0t b ivpu
(
e0 ut st
)
cn b
a N
e r
n t
l eid a1 df d e C(
u i
r i
E iu pt s r
R 2 p
G q
e ece e e n i G.
o r d a a N 70 p
e n
r eb i
M R
o epu p gse o0 5 e
a pS A
n N
v st n
eluer r t 5 t
e t
a c
e l
pb eh nor r
E o
it id r
uht ot a a
T it ed l
,t P
p r
S N
c a.
t o it t f nia G.d feoh anR e
E O
A eap t
icaF s
c d se eS r
c I
u L
T ge ps a%a it rsC a
i f
r E
C ub e G.
e si p0 ou0 d
D E
uh S
Inh e l
t r 2i Np1 e
n Pt t O
P re d
M S
is N
I n
E o
t L
c l
P u
1 2
3 s
s M
e C
C C
2 i
A R
d o
S h
t 1
T e
S m
S 1
T NI r
ia 0
p U
s e
2 e e
iz f b r
S ou h
S t
c A
e md a
X lp ue E
E m
mi T
r ia a
S np H
i e
mr T
U f
Ao O
S
=
I AEACTOR 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 (ltem 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 structuralintegrity of this portion of the RCS will be maintained. The program for inservice j
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 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 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 shutdown and an unscheduled inspection, during which the leaking tubes will be q
located and plugged or (for Model E steam generators only) repaired. Defective tubes in Model E steam generators may be repaired by a Westinghouse laser welded sleeve. The j
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 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. Except as discussed below, plugging or (for Model E steam generators only) 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 con % Westinghouse laser welded sleeve with imperfection exceeding 40% of nominal wall thickt
, it must be plugged. The basis for the sleeve plugging limit for Model E steam generators is based on Regulatory Guide 1.121 analysis, and is described in the Westinghouse sleeving technical reposts listed above. Steam generator tube inspections of operating SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2a Unit 1 - Amendment No. 55;ee;e9;90,107 Unit 2 - Amendment No. 44;77, 94
r REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) 3 plants have demonstrated the capability to reliably detect degradation that has penetrated 20%
of the original tube wall thicknese. Repaired tubes are also included in the inservice tube inspection program.
For Model E steam generators only, the voltage-based repair limits of SR 4.4.5 l
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 criteria of GL 95-05 are also applicable to the Unit 2 flow distribution 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 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 for Model E steam generators requires a derivation of the l
voltage structural limit from the burst versus voltage empirical correlation and tnen the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
1 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; Vunt, is determined from the structural voltage limit by applying the following equation:
Vunt = Vst - Von - Vune where Von represent the allowance for flaw growth between inspections and Vuos 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.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3 Unit 1 - Amendment No. 09;&3;90;96,107 Unit 2 - Amendment No. -79;03,94
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 of Model E steam generators in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements severa! reporting requirements for Model E steam generators l
recommended by GL 95-05 for situations which the NRC wants to be notified prior to retuming 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 vo9 age 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 fallinto 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 j
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 1
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 1
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 additionalleakage.
l SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3a Unit 1 - Amendment No. 09;96,107 Unit 2 - Amendment No. W,94
4 REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)
The leakage limits incorporated into SR 4.4.6 are more restrictive than the standard operating leakage limits and were implemented in conjuction with the application of voltage-based repair criteria and laser-welded sleeving to Model E steam generators. They were intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or unexpectedly extend outside the thickness 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. The additional margin provided by the reduced leakage limit will be retained with the A94 steam generators.
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 values in the event of either a steam generator tube l
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 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of
.I UNIDENTIFlED LEAKAGE by the Laakage 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 Mries check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should 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. 09;90;96,107 Unit 2 - Amendment No. W;09,94