ML20070Q231
| ML20070Q231 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/07/1994 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070Q241 | List: |
| References | |
| NUDOCS 9405130098 | |
| Download: ML20070Q231 (18) | |
Text
.
l A M )' S UNITED STATES li
- i NUCLEAR REGULATORY COMMISSION C (f t
WASHINGTON, D.C. 2055!W)001 y
l
....+
l l
l COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 50 License No. NPF-72 1.
The Nuchu Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated April 25, 1994, as supplemented April 28, 1994, l
April 30,1994, May 2,1994, and May 6,1994, complies with the l
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, 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 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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
l 9405130098 940507 PDit ADOCK 05000456 p
~2-(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 50 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR THE-NUCLEAR REGULATORY COMMISSION W I, W
ames E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 7, 1994 e
ATTACHMENT TO LICENSE AMENDMENT NO. 50 FACILITY OPERATING LICENSE NO. NPF-72 DOCKET NO. STN 50-456 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Paaes Insert Paaes 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-14a 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-17a 3/4 4-17b 3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19 3/4 4-2i 3/4 4-21 3/4 4-27 3/4 4-27 8 3/4 4-3 8 3/4 4-3 B 3/4 4-3a i
B 3/4 4-4 B 3/4 4-4 l
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS 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 steam generator (s) to OPERABLE status prior to increasing T,y 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 Inspection - 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 Inspection - The steam 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 fre-quencies 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 expectations of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by the sleeve are not considered an area requiring reinspection.
The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these insperiions 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:
- When refer,ing to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.5.4.a.10.
I BRAIDWOOD - UNITS 1 & 2 3/4 4-13' AMENDMENT NO. 50
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 1)
All tubes that previously had detectable tube wall penetrations greater than 20% that have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 2)
Tubes in those areas where experience has indicated potential
- problems, 3)
At least 3% of the t'otal number of sleeved tubes in all four steam generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the tube and the sleeve, and 4)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
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 thost 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 I Cycle 5, implementation of the tube support plate interim plugging criteria limit requires a 100% bobbin coil probe inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with outer diameter stress corrosion cracking (005CC) indications.
An inspection using a rotating pancake coil (RPC) probe is required in order to show OPERABILITY of tubes with flaw-like bobbin coil signal amplitudes greater than 1.0 volt but less than or equal to 2.7 volts.
For tubes that will be administratively plugged or repaired, no RPC inspection is required. The RPC results are to be evaluated to establish that the principal indications can be characterized as ODSCC.
The results of each sample inspection shall be classified into one of the following three categories:
Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
BRAIDWOOD - UNITS 1 & 2 3/4 4-14 UNIT 1 -
AMENDMENT N0.50
-BEACTOR COOLANT SYSTEM SURVE1LLANCE REQUIREMENTS (Continued)
C-2 One or mcre 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 degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10% of wall thickness) further wall penetrations to be included in the above percentage calculations.
BRAIDWOOD - UNITS 1 & 2 3/4 4-14a UNIT 1 - AMENDMENT NO. 50
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.3 Jyspection Frecuencies - The above required inservice inspections of 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 initial criticality. Subsequent inservice 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 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 lable 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)
Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2c., or 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 3)
A Condition IV loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 4)
A Condition IV main steam line or feedwater line break.
BRAIDWOOD - UNITS 1 & 2 3/4 4-15 AMENDMENT N0. 50
REACTOR COOLANT SYSTEM SVRVEILLANCE REDUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this specification:
1)
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube or sleeve 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 or sleeve; 3)
Dearaded Tube means a tube or sleeve containing unrepaired imperfections greater than or equal to 20% of the nominal tube or sleeve wall thickness caused by degradation; 4)
% Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation; 5)
Defect means an imperfection of such severity that it exceeds the plugging or repair limit.
A tube or sleeve containing an unrepaired defect is defective; 6)
Pluaaina or Repair 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.
The plugging or repair limit imperfection depth is equal to 40% of the nominal wall thickness.
For Unit 1 Cycle 5, this definition does not apply to the region of the tube subject to the tube support plate interim plugging criteria limit, i.e., the tube support plate intersections.
Specification 4.4.5.4.a.ll describes the repair limit for use within the tube support plate intersection of the tube; 7)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above; 8)-
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
For a tube that has been repaired by. sleeving, the tube inspection shall include the sleeved portion of the tube, and BRAIDWOOD - UNITS 1 & 2 3/4 4-16 UNIT 1 - AMENDMENT NO. 50
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10)
Tube Repair refers to a process that reestablishes tube serviceability. Acceptable tube repairs will be performed by the following processes:
a)
Laser welded sleeving as described by Westinghouse report WCAP-13698, Rev. 1, or b)
Kinetic welded sleeving as described by Babcock & Wilcox Topical Report BAW-2045PA, Rev. 1.
Tube repair includes the removal of plugs that were previously installed as a corrective or preventative measure.
A tube inspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service.
- 11) Tube Support plate Interim Criteria limit is used in Unit 1 for the disposition of a steam generator tube for continued service that is experiencing ODSCC confined within the thickness of the tube support plates.
For application of the tube support plate interim plugging criteria limit, the tube's disposition for continued service will be based upon standard bobbin coil probe signal amplitude of flaw-like indications.
Pending incorporation of the voltage verification requirements in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in Unit I steam generator inspections for consistent voltage normalization.*
1.
A tube can remain in service with a flaw-like bobbin coil
. signal amplitude of less than or equal to 1.0 volt, regardless of the depth of the tube wall penetration, provided Item 3 below is satisfied.
2.
A tube can remain in service with a flaw-like bobbin coil signal amplitude greater than 1.0 volt but less than or equal to 2.7 volts provided an RPC inspection does not detect degradation and provided Item 3 below is satisfied.
- The plant specific guidelines used for all inspections shall be consistent with the eddy current guidelines in Appendix A of WCAP-13854.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17 UNIT 1 - AMENDMENT NO.50
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 3.
The projected distribution of crack indications after 100 calendar days of operation from restart of Unit 1 Cycle 5, not counting any time below a T temperature of 500*F, is 3
verified to result in total prim,a,ry to secondary leakage less than 26 gpm (includes operational and accident leakage). The basis for determining expected leak rates from the projected crack distribution is contained in Attachment A of Ceco's letter dated April 30, 1994.
4.
A tube with a flaw-like bobbin coil signal amplitude of greater than 2.7 volts shall be plugged or repaired.
Certain tubes identified in Westinghouse letter report NSD-TAP-3069, "Braidwood 1: Technical Support for Cycle 5 S/G Interim Plugging Criteria, Pre-WCAP Release," dated April 21, 1994, shall be excluded from application of the tube support plate interim plugging criteria limit.
It has been determined that these tubes may collapse or deform following a postulated LOCA + SSE.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all tubes exceeding the plugging or repair limit) required by Table 4.4-2.
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 ea;n Meam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; 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.
~
Results'of steam generator tube inspections which fall into Category c.
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 investi-gations conducted to determine cause of the tube degradation and corre'ctive measures taken to prevent recurrence.
1 BRAIDWOOD - UNITS 1 & 2 3/4 4-17a UNIT 1 - AMENDMENT NO.50
REACTOR COOLANT SYSTEM SVRVElllANCE RE0VIREMENTS (Continued) d.
For Unit I Cycle 5, the results of inspection for all tubes in which the tube support plate interim plugging criteria limit has been applied shall be reported to the Commission pursuant to Specification, 1
J 6.9.2 following completion of the steam generator tube inservice inspection and prior to Cycle 5 operation. The report shall include:
1.
Listing of the applicable tubes, 2.
Location (applicable intersections per tube) and extent of degradation (voltage), and 3.
Projected Steam Line Break (MSLB) Leakage.
e-BRAIDWOOD - UNITS 1 & 2 3/4 4-17b UNIT 1 - AMENDMENT NO. 50
TABLE 4.4-1 liUjjMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection Yes s
No. of Steam Generators per Unit Four first Inservice Inspection Two Second & Subsequent inservice Inspections One' TABLE NOTATION 1.
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % 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 generaturs 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 circumstances the sample sequence shall be modified to inspect the most severe conditions.
Each 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 subsequent inspections shall follow the instructions described above.
BRAIDWOOD - UNITS 1 & 2 3/4 4-18 AMENDMENT N0. 50
m TARtE 4.4 2 4
tn STEAM GENERATOR TUDE INSPECilON r;
- ,3 l_;
IST SAMPLE INSPFCT!DN 2ND S AMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Site Residt Action Reetuited Restell Action Reepsted -
RestAt Action Required 5
A mininum of C-1 None N.A.
N.A.
N.A.
N.A.
d S Tubes per
-l:
C2 Plug or repair C1 None N.A.
N.A.
I 3, g, delective tubes and
[
C2 Nuq o repair M
Nm l
l inspect additional 2S tubes in tinis d ' I " C ' " ' "I'*
- and inspect C2
% w r@
[
- 5. G' aililitional 45 defective tubes tubes in this S. G.
C-3 Perform actiori foe C-3 result of first i
sa m ple t'
C.3 Perfoem action for N.A.
N.A.
I C-3 result of first
[
sa m ple e
C3 Inspect all tubes in AM other None N.A.
N.A.
this S. G., plug or S. G.s are repair defective C-1 tubes and inspect 25 tubes in each Some S. G.s Perform action for N.A.
N.A.
other S. G.
C-2 but no.
C.2 result of additionel-secoral sample Notification to NRC S. G. ese C-3 fSO.72 thi 21 of 10 8.,'e.c., S. o..r.d
^
'"P**"'
"'A
" A-i S. o. is C.3 3
c,,,,,, so a
plug or repeir l>
3 defective tubes.
E Notification to NRC pursuant to o
150.72(bil2) of 10 CFR Part 50 0 \\
S.= 3"1 lhere N is the number of steam ejenerators in the unit, arul n is the number ^nt sicam 5
n etoneratnes inspected durinet an inspection
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
I gpm UNIDENTIFIED LEAKAGE, c.
600 gallons per day total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 150 gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig, and f.
I gpm leakage at a Reactor Coolant System pressure of 2235 i 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-I.*
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a.
With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, 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 M0T 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 350 psig are allowed.
Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proporational to pressure differential to the one-half power.
BRAIDWOOD - UNITS 1 & 2 3/4 4-21 UNIT 1 - AMENDMENT NO. 50
REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACT1VITY 1
\\
LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, l
a.
and**
b.
Less than or equai to 100/E microcuries per gram of gross radioactivity.
APPLICABILITY:
H0 DES 1, 2, 3, 4, and 5.
&G ION:
T MODES 1, 2 and 3*:
a.
With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,,, less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and b.
With the specific activity of the reactor coolant greater than 100/5 microcuries per gram, be in at least H0T STANDBY with T,y, less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- With Tavg greater than or equal to 500*F.
- For Unit 1 Cycle 5, the steam generators will be considered OPERABLE for the first 100 calendar days of operation with T greater than 500*F.
During that time, reactor coolant DOSE EQUIVALENT Y*131 will be limited to 0.35 microCuries per gram.
BRAIDWOOD - UNITS 1 & 2 3/4 4-27 UNIT 1 - AMEN'JMENT NO. 50 e
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspect' ion 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 surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage - 150 gallons per day per steam generator).
l Cracks having a reactor-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 reactor-to-secondary leakage of 150 gallons per day per l
steam generator can readily be detected by radiation monitors of steam generator blowdown.
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 by sleeving. The technical bases for sleeving are described in Westinghouse report WCAP-13698 Rev. I and Babcock & Wilcox Topical Report BAW-2045PA Rev. 1.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the plugging or repair limit of 40% of the tube nominal wall thickness.
If a sleeved tube is found to contain a through wall penetration in the sleeve of equal to or greater than 40% of the nominal wall thickness, the tube must be plugged. The 40% plugging limit for the sleeve is derived from Reg. Guide 1.121 analysis and utilizes a 20% allowance for eddy current uncertainty and additional degradation growth.
Inservice inspection of sleeves is required to ensure RCS integrity.
Sleeve inspection techniques are described in Westinghouse Report WCAP-13698 Rev. I and Babcock & Wilcox Tupical Report BAu-2045PA Rev. 1.
Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation that has penetrated 20% of the pressure' retaining portions of the tube or sleeve wall thickness.
Commonwealth Edison will validate the adequacy of any system that is used for periodic inservice inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 UNIT 1 - AMENDMENT NO. 50
REACTOR'C00LANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (continued)
For Unit 1 Cycle 5, tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates will be dispositioned in accordance with Specification 4.4.5.4.a.11.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 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.
f I
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3a UNIT 1 - AMENDMENT N0. 50
REACTOR COOLANT SYST 8 BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE B0UNDARY 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.
The total steam generator tube leakage limit of 600 gpd for all steam l
generators 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 rupture or steam line break.
The 600 gpd limit is consistent with 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 UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This li,itation ensures that in the event of a LOCA, the Safety injection flow will not be less than assumed in the safety analyses.
The 1 gpm 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 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, those valves should be tested periodically to ensure low-probability of gross failure.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-4 UNIT 1 - AMENDMENT N0. 50
_