ML20248L871
| ML20248L871 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/12/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20248L873 | List: |
| References | |
| NUDOCS 9803240323 | |
| Download: ML20248L871 (20) | |
Text
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'4 UNITED STATES j
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30eeH001
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PACIFIC GAS AND FI FCTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCI FAR POWER PLANT. UNIT NO.1 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.124 License No. DPR-80 1.
The Nuclear Regulatory Commission (the Commission) has found that:
1 A.
The application for amendment by Pacific Gas and Electric Company (the
{
licensee) dated February 26,1997, as supplemented by letters dated December 23,1997, January 30,1998, and February 9,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 satisfM.
1 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) of Facility Operating License No. DPR-80 is hereby amended to read as follows:
l l
9003240323 980312 PDR ADOCK 05000275 P
6
. I (2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.124., are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance to be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
1
% W
^
Steven D. Bloom, Project Manager Project Directorate IV-2 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications l
Date of issuance:
March 12, 1998 l
l
p QCtg O
UNITED STATES j
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20066-0001 49.....,o PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO dANYON NUCL FAR POWER PLANT. UNIT NO. 2 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.122 License No. DPR-82 I
1.
The Nuclear Regulatory Commission (the Commission) has found that:
I A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated February 26,1997, as supplemented by letter dated December 23,1997, January 30,1998, and February 9,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the j
Act), and the Commission's regulations set forth in 10 CFR Chapter 1, 1
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; l
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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as fc,Ilows:
l
i b
l l.
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.122, are hereby incorporated in the license. Pacific Gas and Electric i
Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance to be implemented within 30 days of the date ofissuance.
FOR THE NUCLEAR REGULATORY COMMISSION aoi Da teven D. Bloom, Project Manager Project Directorate IV-2 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
March 12, 1998 l
.b ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO.124 TO FACILITY OPERATING LICENSE NO. DPR-80 l
AND AMENDMENT NO.122 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 1
Revise Appendix A Technical Specifications by removing the pages identified below and l
inserting the enclosed pages. The revised pages are identified by Amendment number and contain marginallines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 4-12 3/4 4-12 3/4 4-14 3/4 4-14 3/4 4-14a 3/4 4-14b 3/4 4-15 3/4 4-15 3/4 4-19 3/4 4-19 3/4 4-20 3/4 4-20 B 3/4 4-3 B 3/4 4-3 B 3/4 4-3a B 3/4 44 B 3/4 44 B 3/4 4-5 B 3/4 4-5 l
l l
l 1
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 generator (s) to OPERABLE status prior to increasing T,yg 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 requirement of Specification 4.0.5.
4.4.5.1 Steam Generator Sample 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 Sample 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 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.
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 j
these inspections shall be selected on a random basis except:
I Where experience in similar plants with similar water a.
I 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 l
generator shall include:
1)
All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
1 DIABLO CANYON - UNITS 1 & 2 3/4 4-11
SURVE)LLANCE REQUIREMENTS (Continued) 2)
Tubes in those areas where experience has indicated potential problems, and 3)
A tube inspection ()ursuant to S) edification 4.4.5.4a.8) shall be performed on eac1 selected tuae.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
4)
Indications left in service as a result of application of the tube support ) late voltage-based repair criteria shall be inspected by.)obbin coil probe during all future refueling outages.
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1)
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and 2)
The inspections include those portions of the tubes where imperfections were previously found.
d.
Implementation of the steam generator tube / tube support plate repair criteria requires a 100% bobbin coil inspection for hot-leg and cold-leg tube support alate intersections down to the lowest cold-leg tube support plate witi known outside diameter stress corrosion cracking (ODSCC) indications.
The determination or the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20% 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:
Cateoory InsDection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes. but not more than 1% of the total tubes inspected are defective, or between 5% and 10%
of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are 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.
DIABLO CANYON - UNITS 1 & 2 3/4 4-12 Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - 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 crit-icality. 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 addi-tional degradation has occurred, the inspection interval may be ex-tended to a maximum of once per 40 months; b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals 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 exter.ded 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 Speci-fication 3.4.6.2; or 2)
A seismic occurrence greater than the Double Design 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.
l DIABLO CANYON - UNITS 1 & 2 3/4 4-13
i.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 AcceDtance Criteria a.
As used in this Specification:
1)
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or s )ecifications.
Eddy-current testing indications below 20% of tw nominal tube wall thickness, if detectable, may be considered as imperfections:
2)
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube:
3)
Dearaded T_yhg means a tube containing imperfections greater than or equal to 20% of the nominal wall thiccness caused by degradation:
4)
% Degradation means the percentage of the tube wall thickness affected or removed by degradation:
5)
Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective:
6)
Pluaaina limit means the' imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness. This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.
Refer to 4.4.5.4a.10) for the repair limit applicable to these intersections:
7)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of a Double Design Earthquake, a loss-of-coolant accident or a steam line or feedwater line break as specified in 4.4.5.3c., above:
8)
Tube Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) com U-bend to the top support of the cold leg:pletely around the and 9)
Preservice Insoection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be aerformed after the field hydrostatic test and prior to initial 30WER OPERATION using the equipment and techniques expected to be used during subsequent i
inservice inspections.
DIABLO CANYON - UNITS 1 & 2 3/4 4-14 Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 10) Tube SuoDort Plate Pluaaino Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging limit is based on maintaining steam generator tube serviceability as described 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 voltages less than or equal to the lower voltage repair limit [ Note 1].
will be allowed to remain in service.
b.
Steam generator tubes. whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit [ Note 1]. will be repaired or plugged except as noted in 4.4.5.4a.10)c below.
c.
Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit
[ Note 1] but less than or equal to the upper voltage repair limit [ Note 2]. may remain in service if a rotating pancake d
coil inspection does not detect degradation.
Steam generator tubes. with indications of outside diameter l
stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit [ Note 2] will be plugged or repaired.
d.
Certain intersections as identified in Westinghouse letter to PG&E dated Se)tember 3.1992. " Deformation of Steam Generator Tubes r llowing a Postulated LOCA and SSE Event."
o will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA + SSE event.
e.
If an unscheduled mid-cycle inspection is performed. the following mid-cycle repair limits apply instead of the limits identified in 4.4.5.4a.10)a. 4.4.5.4a.10)b. and 4.4.5.4a.10)c. The mid-cycle repair limits are determined from the following equations:
DIABLO CANYON - UNITS 1 & 2 3/4 4-14a Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
)
REACTOR COOLANT SYSTEM SURVEIll ANCE RE0VIREMENTS (Continued)
V"'=
V 1.0 + NDE E Gr ( CL-At
)
CL V,t,t - V,o - (V,t - V,t) ( CL-At )
t l
CL where:
V,t
- upper vpltage repair limit V
- lower voltage repair limit V,t,t
- mid-cycle upper voltage repair limit based on t
time into cycle V
,t
- mid-cycle lower voltage repair limit based on g
V and time into cycle At
- lEgth of time since last scheduled inspection during which V and V were implemented CL
= cycle length (Die time btetween two scheduled t
t steam generator inspections)
VSL
= structural limit voltage Gr
- average growth rate 3er cycle length NDE
- 95% cumulative proba)ility allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC)
Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4a.10)a.
4.4.5.4a.10)b. and 4.4.5.4a.10)c.
Note 1:
The lower voltage repair limit is 2.0 volts for 7/8-inch diameter tubing at DCPP Units 1 and 2.
Note 2:
The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.
b.
The steam generator shall be determined OPERABLE after completirg the corresponding actions (plug all tubes exceeding the plugging limit) l required by Table 4.4-2.
l DIABLO CANYON - UNITS 1 & 2 3/4 4-14b Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
]
SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.5 Reoorts a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam 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 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.
c.
Results of steam generator tube inspections, which fall into Category C-3. shall be reported in a Special Report to the Commission l
pursuant to Specification 6.9.2 within 30 days and prior to resump-tion 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 implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC prior to returning the steam generators to service should any of the following conditions i
arise:
1)
If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit determined from the licensing basis dose calculation for the postulated main steamline break for the next operating cycle.
l 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 10'2.
notify the NRC and provide an assessment of the safety significance of the occurrence.
DIABLO CANYON - UNITS 1 & 2 3/4 4-15 Unit 1 - Amendment No.124 Unit 2 - Amendment No.122 J
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
l a.
b.
1 gpm UNIDENTIFIED LEAKAGE.
c.
150 gallons per day of primary-to-secondary leakage through any one steam generator.
l d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig, and f.
1 gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig for Reactor Coolant System Pressure Isolation Valves as specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2. 3*, and 4*.
l ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE. be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above iimits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System pressure isolation valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual and/or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- For Modes 3 and 4. if steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 microcuries/ml for princiaal gamma emitters, the leakage requirement of Specification 3.4.6.2c may )e considered met.
I DIABLO CANYON - UNITS 1 & 2 3/4 4-19 Unit 1 - Amendment No. M.124 Unit 2 - Amendment No. E,122
~.
SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a.
Monitoring the containment atmosphere particulate or gaseous radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; b.
Monitoring the containment structure sump inventory and discharge at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
c.
Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals at least once per 31 days when the Reactor Coolant System pressure is 2235 20 psig with the modulating valve fully open.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; d.
Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, except when T is being changed by greater than 5'F/ hour or when diverting
- reactor coolant to the liquid holdup tank, in which cases the required inventory balance shall be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of the excepted operation; e.
' Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and f.
Determination of steam generator primary-to-secondary leakage at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.4.6.2.2 As specified in Table 3.4-1 Reactor Coolant System pressure isola-tion valves shall be demonstrated OPERABLE pursuant to Specification 4.0.5.
except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:
a.
At least once each REFUELING INTERVAL during startup, b.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and c.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve. After each disturbance of the valve, in lieu of measuring leak rate, leak-tight integrity may be verified by absence of pressure buildup in the test line downstream of the valve.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4 DIABLO CANYON - UNITS 1 & 2 3/4 4-20 Unit 1 - Amendment No, %.-1g)24 Unit 2 - Amendment No. y g }22
REACTOR COOLANT SYSTEM BASES I
STEAM GENERATORS (Continued) 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 (primary-to-secondary leakage = 150 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
DCPP has demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected during power operation.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
The voltage-based repair limits of SR 4.4.5.4a.10) 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 DDSCC that occurs at other locations within the SG. Additionally, the l
repair criteria ap)1y only to indications where the degradation mechanism is I
dominantly axial 0) SCC with no significant cracks extending outside the i
thickness of the support late.
Refer to GL 95-05 for additional description l
of the degradation morpho ogy.
Implementation of SR 4.4.5.4a.10) requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650*F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit, V is determined from the structural voltage limit by applying the folfow,ing equation:
V
-V
- V, - V,,,
n DIABLO CANYON - UNITS 1 & 2 B 3/4 4-3 Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) where V,, represents the allowance for flaw growth between inspections and V representstheallowanceforpotentialsourcesoferrorinthemeasurementoT the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
The mid-cycle equation in SR 4.4.5.4a.10)e. should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5d 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 purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected E0C voltage distributions prior to returning the SGs to service. Note that if leakage and conditional burst probability were calculated using the measured E0C voltage distribution for the purposes of addressing the GL Section 6.a.1 and 6.a.3 re)orting criteria, then the results of the projected EOC voltage distribution s1ould be provided per the GL Section 6.b(c) criteria.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However. even if a wastage defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit defined in Surveillance Requirement 4.4.5.4a.
l Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3. these results will be reported to the Commission as a Special Report pursuant to Specification 6.9.2 within 30 days and arior to resumption of plant o)eration. Such cases will be considered by t1e Commission on a case-)y-case basis and may result in a requirement for i
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 l
l 3rovided to monitor and detect leakage from the reactor coolant pressure youndary. These Detection Systems are functionally consistent with the recommendations of Regulatory Guide 1.45. " Reactor Coolant Pressure Boundary Leakage Detection Systems.".May 1973.
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-3a Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
, i BASES 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 i
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 I
to a threshold value of less than 1 gpm. This threshold value is sufficiently I
low to ensure early detection of additional leakage.
The total steam generator tube leakage limit of 150 gpd for any one steam generator 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.
This i
limit is consistent with the assumptions used in the analysis of these accidents.
The 150 gpd leakage limit per steam generator ensures that steam I
generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. The primary-to-secondary operational leakage limit of 150 g)d per steam generator is more restrictive than the standard operating leacage limits and is 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. 1t will be detected. and the plant shut down in a timely manner.
Calculations for primary to-secondary leakage are performed using' approximate Standard Reference State of 25'C. When determining primary-to-secondary leakage of 150 gpd. indeterminent inaccuracies associated with determination of leakage are not considered.
For Modes 3 and 4. the primary system radioactivity level (source term) may be very low. making it difficult to measure primary-to-secondary leakage of 150 gallons per day.
Therefore. if steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 microcuries/ml for principal gama emitters, the 150 gallons per day leakage limit may be considered met.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA. the safety injection flow will not be less than assumed in the safety analyses.
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-4 Unit 1 - Amendment No.124 Unit 2 - Amendment No.122
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued) l 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 3reventing overpressurization and rupture of the ECCS low pressure piaing whic1 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 consecuent intersystem LOCA.
Leakage from the RCS pressure isolation valves is IDEhTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure l
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 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These j
values are conservative in that specific site parameters of the Diablo Canyon site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
l The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE. EQUIVALENT I-131 but within the allowable limit shown on Figure 3.4-1, accommodates aossible iodine spiking phenomenon which may occur i
following changes in THERMA. POWER. Operation with specific activity levels exceeding 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 should be limited since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.
4 DIABLO CANYON - UNITS 1 & 2 B 3/4 4-5 Unit 1 - Amendment No. & 98.124 Unit 1 - Amendment No. W 97.122
BASES SPECIFIC ACTIVITY (Continued) i The sample analysis for determining the gross specific activity and I can exclude the radiciodines because of the low reactor coolant limit of 1 microcurie /
gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radio-iodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If the gross specific activity level and radioicdine level in the reactor coolant were at their limits, the radiciodine contribution would be approximately IL In a release of reactor coolant with a typical mixture of radioactivity, the actual radiciodine contri-bution would probably be about 20L The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has been made for several reasons.
The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, trans-port, 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 relat-able to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity.
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 radionuclides above and below a half-lif of 10 minutes.
For these reasons the radionuclides that are excluded from con-sideration are expected to decay to very low levels before they could be trans-ported from the reactor coolant to the SITE BOUNDARY under any accident condi-tion.
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 per-form the sampling, transport the sample, and perform 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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. Alternatively, gamma spectroscopy may be used.
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.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in fre-quency of isotopic analyses following power changes may be permissible if justified by the data obtained.
DIABLO CANYON - UNITS 1 & 2 8 3/4 4-6
.