ML20216B829
| ML20216B829 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/14/1998 |
| From: | Hansen A NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216B834 | List: |
| References | |
| NUDOCS 9805180464 | |
| Download: ML20216B829 (16) | |
Text
_ _ _ - _ _ _ _
anau gw
-> t UNITED STATES g
j NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 20566-4001 TOLED0 EDIS0N COMPANY CENTERIOR SERVICE COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET N0. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 220 License No. NPF-3 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Toledo Edison Company, Centerior Service Company, and The Cleveland Electric Illuminating Company (the licensees) dated February 26, 1998, as supplemented by letter dated March 20, 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, 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 Specifications as indicated in the t'!achment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows-9805180464 980414 PDR ADOCK 05000346 P
- (2)
Technical Soecifications The Technical Specifications contained in Appendix A, as revised
/
through Amendment No. 220, are hereby incorporated in the license.
The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented not later than 120 days after issuance.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION Allen G. Hansen, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: April 14, 1998
j ATTACHMENT TO LICENSE AMENDMENT N0, 220 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 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 Insert TS 3/4 4-6b TS 3/4 4-6b TS 3/4 4-8 TS 3/4 4-8 TS 3/4 4-9 TS 3/4 4-9 TS 3/4 4-9a TS 3/4 4-9a TS 3/4 4-10 TS 3/4 4-10 TS 3/4 4-10a TS 3/4 4-10a TS 3/4 4-12 TS 3/4 4-12 TS 3/4 4-15 TS 3/4 4-15 TS 3/4 4-16 TS 3/4 4-16 TS B 3/4 4-2 TS B 3/4 4-2
.TS B 3/4 4-3 TS B 3/4 4-3 TS B 3/4 4-3a TS B 3/4 4-3a TS B 3/4 4-4 TS B 3/4 4-4
REACTOR COOLANT SYSTEM STEAM GENERATORS 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 Insoection - The steam f
generator tube minimum sample size, inspection result classification, and the -
j 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 these insper.tions shall be selected on a random basis except:
a.
The first sample inspection during each inservice inspection of.
l each steam generator shall include:
1.
All tubes or tube sleeves that previously had detectable wall penetrations (> 20%) that have not been plugged or repaired by repair roll or sleeving in the affected area.
(Tubes repaired by sleeving or repair roll remain available for random selection).
2.
At least 50% of the tubes inspected shall be in those areas where experience has indicated potential problems.
. DAVIS-BESSE - UNIT 1 3/4 4-6b Amendment No.+9Br 220
)
SVRVEILLANCE RE0VIREMENTS (Continued)
Notes:
(1)
In all inspections, previously degraded tubes must exhibit significant (> 10%) further wall penetrations to be included in the above percentage calculations.
(2)
Where special inspections are performed pursuant to 4.4.5.2.b, defective or degraded tubes found as a result of the inspection shall be included in determining the Inspection Results Category for that special inspection but need not be included in determining the Inspection Results Category for the general steam generator inspection.
4.4.5.3 Inspection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. Inservice inspections shall be performed at intervals of not less than l
12 nor more than 24 calendar months after the previous inspection.
If the results of two consecutive inspections for a given group
- of tubes following service under all volatile treatment (AVT) conditions fall 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 for that group may be extended to a maximum of 40 months.
- b. If the results of the inservice inspection of a steam generator performed in accordance with Table 4.4-2 at 40 month intervals for a given group
- of tubes fall in Category C-3, subsequent inservice inspections shall be performed at intervals of not less than 10 nor more than 20 calendar months after the previous inspection.
The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in 4.4.5.3a and the interval can be extended to 40 j
months.
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified j
in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1. Primary-to-secondary tube leaks (not including leaks originating from j
tube-to tube sheet wolds) in excess of the limits of Specification 3.4.6.2.
1 If the leak is determined to be from a repair roll joint, rather than j
selecting a random sample, inspect 100% of the repair roll joints in j
the affected steam generator.
If the results of this inspection fall into the C-3 category, perform additional inspections of the new roll
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areas in the unaffected steam generator.
1
- A group of tubes means:
(a)
M1 tubes inspected pursuant to 4.4.5.2.b, or (b)
All tubes in a steam generator less those inspected pursuant to j
4.4.5.2.b.
DAVIS-BESSE, UNIT 1 3/4 4-8 Amendment No. Gh-220
)
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 2. A seismic occurrence greater than the Operating Basis Earthquake.
- 3. A loss-of-coolant accident requiring actuation of the engineered safeguards.
- 4. A main steam line or feedwater line break.
- d. The provisions of Specification 4.0.2 are not applicable.
4.4.5.4 Acceptance Criteria
- a. As used in this Specification:
- 1. Tubino or Tube means that portion of the tube or tube sleeve which forms the primary system to secondary system boundary,
- 2. Imperfection 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.
j
- 3.
Dearadation means a service-induced cracking,
wastage, wear or-general corrosion occurring on either inside or outside of a tube.
j
- 4. Dearaded Tube means a tube containing imperfections ;t 20% of the nominal wall thickness caused by degradation that has not been repaired by repair roll or sleeving in the affected area.
l
- 5. % Dearadation means the percentage of the tube wall thickness affected or removed by degradation.
- 6. Defect means an imperfection of such severity that it exceeds the repair limit.
A defective tube is a tube containing a defect that has not been repaired by repair roll or sleeving in the affected
-l area or a sleeved tube that has a defect in the sleeve.
- 7. Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by repair roll or sleeving in the affected area because it may become l-unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.
The Babcock and Wilcox process described in Topical Report BAW-2120P will be used for sleeving.
DAVIS-BESSE, UNIT 1 3/4 4-9 Amendment No. 21,171,220 i
I
SURVEILLANCE REQUIREMENTS (Continued)
(Continued) 7. The repair roll process will only be used to repair tubes with defects in the upper tubesheet area. The repair roll process will be performed only once per steam generator tube using a 1 inch reroll length.
The new roll area must be free of degradation in order for the repair to be considered acceptable.
The repair roll process used is described in the Topical Report BAW-2303P, Revision 3.
- 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 4.4.5.3.c, above.
- 9. Tube Insoection means an inspection of the steam generator tube from the point of entry completely to the point of exit. The previously existing tube and tube roll, above the new roll area in the upper tube sheet, can be excluded from future periodic inspection requirements because it is no longer part of the pressure boundary once the repair roll is installed.
DAVIS-BESSE, UNIT 1 3/4 4-9a Amendment No. 21,171, 220 l l
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SURVEILLANCE REQUIREMENTS (Continued) 10.
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 performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during st.bsequent inservice inspections,
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug _ or repair by repair roll or sleeving in l
the affected areas all tubes exceeding the repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.
j 4.4.5.5 Reoorts
- a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
- b. The complete results of the steam generator tube inservice inspection
)
shall be submitted on an annual basis in a report for the period in which this-inspection was completed. This report shall include:
j
- 1. Number and extent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection.
- 3. Identification of tubes plugged, sleeved or repair rolled.
l 1
- c. Results of steam generator tube inspections which fall into Category C-3 and require notification of the Commission shall be reported prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
4.4.5.6 The steam generator shall be demonstrated OPERABLE by verifying steam generator level to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.5.7 When steam generator tube inspection is performed as per Section 4.4.5.2, an additional but totally separate inspection shall be performed on special interest peripheral tubes in the vicinity of the secured internal auxiliary feedwater header.
This testing shall only be required on the steam generator selected for inspection, and the test shall require inspection only between DAVIS-BESSE, UNIT 1 3/4 4-10 Amendment No. 0,27,52,171,10', 220
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) the upper tube sheet and the 15th tube support plate. The tubes selected for inspection shall represent the entire circumference of the steam generator and J
shall total at least 150 peripheral tubes.
j 4.4.5.8 Visual inspections of the secured internal auxiliary feedwater header, header to shroud attachment welds, and the external header thermal sleeves shall be performed on each steam generator through the auxiliary I
feedwater injection penetrations.
]
These inspections shall be performed during the third and fourth refueling outages and at the ten-year ISI.
4.4.5.9 When steam generator tube inspection is performed as per Section 4.4.5.2, an additional but totally separate inspection shall be performed on special interest tubes that have been repaired by the repair roll process.
This inspection shall be performed on 100% of the tubes that have been repaired by the repair roll process. The inspection shall be limited to the repair roll joint and the roll transitions of the repair roll.
Defective or degraded tubes found in the repair roll region as a result of the inspection need not be included in determining the Inspection Results Category for the general steam generator inspection.
j DAVIS-BESSE, UNIT 1 3/4 4-10a Amendment No. 6h 220
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, c.
150 GPD primary-to-secondary leakage through the tubes of any one l
steam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
10 GPM CONTROLLED LEAKAGE, and f.
5 GPM leakage from any Reactor Coolant System Pressure Isolation Valve as specified in Table 3.4-2.
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 B0UNDARY LEAKAGE, reduce the leakage i
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 j
hours except as permitted by paragraph c below.
{
c.
In the event that integrity of any pressure isolation valve specified in Table 3.4-2 cannot be demonsta ated, POWER OPERATION may continue, provided that at least two valves in each high pressure line having a j
non-functional valve are in and remain in, the mode corresponding to l
the isolated condition.(a) d.
The provisions of Section 3.0.4 are not applicable for entry into j
MODES 3 and 4 for the purpose of testing the isolation valves in Table 3.4-2.
Motor operated valves shall be placed in the closed position and power supplies deenergized.
DAVIS-BESSE, UNIT 1 3/4 4-15 Geder dtd. 4/20/M Amendment No. 135,100220
)
f i
]
SURVEILLANCE RE0VIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a.
Monitoring the containment atmosphere gaseous or particulate radioactivity 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 sump level and flow indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l c.
Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump l
seals to the makeup system when the Reactor Coolant System pressure
)
is 2185 i 20 psig at least once per 31 days, j
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 /> during steady state operation.
]
e.
An evaluation of secondary water radiochemistry for determination of primary to secondary leakage through the steam generators at least i
once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operations.
1 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in I
Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:
a.
After each refueling outage, b.
Whenever the plant has been in COLD SHUTDOWN for 7 days, or more, j
and if leakage testing has not been performed in the previous 9 months, and c.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve, d.
The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.
4.4.6.2.3 Whenever the integrity of a pressure isolation valve listed in Table 5.4-2 cannot be demonstrated, determine and record the integrity of the high pressure flowpath on a daily basis.
Integrity shall be determined by performing either a leakage test of the remaining pressure isolation valve, or a combined leakage test of the remaining pressure isolation valve in a series with the closed motor operated containment isolation valve.
In addition, record the position of the closed motor-operated containment isolation valve located in the high pressure piping on a daily basis.
&dee-dated 4/20/01 DAVIS-BESSE, UNIT 1 3/4 4-16 Amendment No. F,135,MO,105, 220
REACTOR COOLANT SYSTEM FASES
}/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that 'the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and pilot operated relief valve against water relief.
The low level limit is based on providing enough water volume to prevent a reactor coolant system low pressure condition that would actuate the' Reactor Protection System or the Safety Feature Actuation System.
The high level limit is based on providing enough steam volume to prevent a pressurizer high level as a result of any transient.
The pilot operated relief valve and steam bubble. function to relieve RCS pressure during all design transients. Operation'of the pilot operated relief valve minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
3/4.4.5 STEAM CENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surve~illance 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. A process equivalent to the inspection method described in Topical Report BAW-2120P will be used for inservice inspection of steam generator tube sleeves.
This inspection will provide ensurance of RCS integrity.
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 chemistry 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 primary coolant system and the secondary coolant system (primary-to-secondary leakage =
150 GPD through any one steam generator).- Cracks having a primary-to-l secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal DAVIS-BESSE, UNIT 1 B 3/4 4-2 Amendment No. 135,171,220
.e i
REACTOR COOLANT SYST@
BASES (Continued)
{
operation and by postulated accidents. Operating plants have demonstrated that
)
primary-to-secondary leakage of 150 GPD can be detected by monitoring the l
secondary coolant.
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 repair rolling or sleeving in the affected l
areas.
Wattage-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.
As described in Topical Report BAW-2120P, degradation as small as 20% through wall can be detected in all areas of a tube sleeve except for the roll expanded areas and the sleeve end, where the limit of detectability is 40%
through wall.
Tubes with imperfections exceeding the repair limit of 40% of the nominal wall thickness will be plugged or repaired by repair rolling or l
sleeving the affected areas. Davis-Besse will evaluate, and as appropriate implement, better testing methods which are developed and validated for commercial use so as to enable detection of degradation as small as 20%
l through wall without exception.
Until such time as 20% penetration can be detected in the roll expanded areas and the sleeve end, inspection results will be compared to those obtained during the baseline sleeved tube inspection.
i An additional repair method for degraded steam generator tubes consists of rerolling the tubes in the upper tubesheet to create a new roll area and pressure boundary for the tube.
The repair roll process will ensure that the area of degradation will not serve as a pressure boundary, thus permitting the tube to remain in service.
The degraded area of the tube can be excluded from future periodic inspection requirements because it is no longer part of the pressure boundary once the repair roll is installed in the upper tubesheet.
All tubes which have been repaired using the repair roll process will have the new roll area inspected during the inservice inspection.
Defective or degraded tube indications found in the new roll area as a result of the inspectio: of the repair roll and any indications found in the originally rolled region of the rerolled tube need not be included in determining the Inspection Results Category for the general steam generator inspection.
The repair roll process will be performed only once per steam generator tube using a 1 inch reroll length as described in the Topical Report BAW-2303P, Revision 3.
Thus, multiple applications of the rerolling process to any individual tube is not acceptable.
The new roll area must be free of degradation in order for the repair to be considered acceptable. After the new roll area is initially deemed acceptable, future degradation in the new roll area will be analyzed to determine if the tube is defective and needs to be removed from service. The rerolling process is described in the Topical Report BAW-2303P, Revision 3.
DAVIS-BESSE, UNIT I B 3/4 4-3 Amendment No. 171,184,4 % 220
I REACTOR COOLANT SYSTEM BASES (Continued)
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results shall be reported to the Commission 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.
The steam generator water level limits are consistent with the initial assumptions in the USAR.
While in MODE 3, examples of Main Feedwater Pumps l
that are incapable of supplying feedwater to the Steam Generators are tripped pumps or a manual valve closed in.the discharge flowpath. The reactivity requirements to ensure adequate SHl!TDOWN MARGIN are provided in plant l
operating procedures.
l l
l DAVIS-BESSE, UNIT I B 3/4 4-3a Amendment No. 171,184,192, 220 l
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BASES l
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE I
l 3/4.4 /.1 LEAKAGE DETECTION SYSTEMS j
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The RCS-leakage detection systems required by this specification are l
provided to detect and monitor leakage from the Reactor Coolant Pressure Boundary.= These detection systems.are consistent with the recommendation of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE l
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 reouires the unit to be promptly placed in COLD SHUTDOWN.
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Industry experience has shown ti..t, while a limited amount of leakage is 3
expected from the RCS, the UNIDENTIFIED LEAKAGE portion of this can be reduced j
to a threshold value of less that 1 GPM. This threshold value is sufficiently 1
low to ensure early detection of additional leakage.
i The steam generator tube leakage limit of 150 GPD through any one steam generator ensures that the dosage contribution from tube leakage will be limited to a small fraction of 10 CFR Part 100 limits in the event of either a steam generator tube rupture or steam line break.
A 1 GPM total primary to secondary leakage limit is used in the analysis of these accidents.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a i
limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
4 The CONTROLLED LEAKAGE limit of 10 GPM restricts operation with a total RCS leakage from all RC pump seals in excess of 10 GPM.
1 The surveillance requirements for RCS Pressure Isolation Valves provide I
added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
DAVIS-BESSE, UNIT 1 B 3/4 4-4 Amendment No. 4807 220