ML20101L645
| ML20101L645 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/01/1996 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duquesne Light Co, Ohio Edison Co, Pennsylvania Power & Light Co |
| Shared Package | |
| ML20101L650 | List: |
| References | |
| GL-95-005, DPR-66-A-198 NUDOCS 9604040140 | |
| Download: ML20101L645 (16) | |
Text
____..
p uau 24 UNITED STATES y
B NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 206MH)001
- f
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DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.198 License No. DPR-66 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Duquesne Light Company, et al. (the licensee) dated December 7, 1995, as supplemented January 4, March 1, March 5, March 7, March 11, March 27, and March 29, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission'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.
ll 9604040140 960401 I
PDR ADOCK 05000334 P
. 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-66 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.198, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance, to be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION M
I I
J hn
. Stolz, Direc M r Pr ect Directorate G 2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 1, 1996 l
ATTACHMENT TO LICENSE AMENDMENT N0. 198 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 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 areas of change.
Remove Insert 3/4 4-9 3/4 4-9 3/4 4-10b 3/4 4-10b 3/4 4-10c 3/4 4-10c 3/4 4-10d 3/4 4-10d 3/4 4-10e 3/4 4-10e 3/4 4-10f 3/4 4-10f 3/4 4-10g 3/4 4-10h 3/4 4-13 3/4 4-13 8 3/4 4-2a B 3/4 4-2a B 3/4 4-2b
)
B 3/4 4-3g B 3/4 4-39 B 3/4 4-3h B 3/4 4-3h 4
.DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2.
Tubes in those areas where-experience has indicated potential problems, and 3.
At least 3 percent of the total number of sleeved tubes in all three steam generators.
A sample size less than 3 percent is acceptable provided all the sleeved tubes in the steam generator (s) examined during the refueling outage are inspected.
These inspections will include both the tube and the sleeve, and 4.
A tube inspection pursuant to Specification 4.4.5.4.a.8.
If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
5.
Indications left in service as a
result of application of the tube support plate voltage-based repair criteria (4.4.5.4.a.10) shall be inspected by bobbin coil probe during all future refueling outages.
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
i 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 l
where imperfections were previously found.
l l
d.
Implementation of the steam generator tube-to-tube j
support plate repair criteria requires a
100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg l
tube support plate with known outside diameter stress l
corrosion cracking (ODSCC) indications.
The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.
The results of each sample inspection shall be classified into one j
of the following three categories:
BEAVER VALLEY - UNIT 1 3/4 4-9 Amendment No.198
..- -.... _.. ~ _ =
,DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 6.
Pluaaina or Renair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in i
the affected area because it may become unserviceable prior to the next inspection.
The plugging or repair limit imperfection depths are specified in percentage of nominal wall thickness as follows:
a.
Original tube wall 40%
\\
This definition does not apply to tube support j
plate intersections for which the voltage-based repair criteria are being applied.
Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.
b.
Babcock & Wilcox kinetic welded sleeve wall 40%
c.
Westinghouse laser welded sleeve wall 25%
1 7.
Unserviceable describes the condition of a tube if s
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 steamline or feedwater line break as l
specified in 4.4.5.3.c, above.
8.
Tube Insoection means an inspection of the steam generator tube from the point of entry (hot-leg l
side) completely around the U-bend to the top support to the cold-leg.
l 9.
Tube Repair refers to sleeving which is used to maintain a tube in-service or return a tube to service.
This includes the removal of plugs that were installed as a
corrective or preventive
)
measure.
The following sleeve designs have been i
found acceptable:
a.
Babcock Wilcox kinetic welded
" tooling" and installation process parameter changes.
b.
Westinghouse laser welded sleeves, WCAP-13483, Revision 1.
BEAVER VALLEY - UNIT 1 3/4 4-10b Amendment No.1 P
.DPR-66 REACTOR COOLANT SYSTEM i
SURVEILLANCE REQUIREMENTS (Continued) 10.
Tube Succort Plate Pluacina 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 (repair) 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 2.0 volts 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 2.0 volts will be repaired or plugged, except as noted in 4.4.5.4.a.10.c below.
c.
- tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin j
voltagegreaterthan2.0voltsbutlesstganor 8
equal to the upper voltage repair _ limit may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair
)
limit will be plugged or repaired.
d.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.
(1) The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.
BEAVER VALLEY - UNIT 1 3/4 4-10c Amendment No.1"
,DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
The mid-cycle repair limits are determined from the following equations:
Vmmt =
' CL - A t' 1.0 + NDE + Gr CL >
Vwn = Vuun-(Vun-hn) CL-At*
< CL >
where:
upper voltage repair Vma
=
limit lower voltage repair Vum
=
limit mid-cycle upper voltage V,,3
=
repair limit based on time into cycle mid-cycle lower voltage Vmat
=
repair limit based on Vams and time into cycle At length of time since
=
last scheduled inspection during which Vme and Vue were implemented cycle length (the time CL
=
between two scheduled steam generator inspections) structural limit voltage Vst
=
average growth rate per Gr
=
cycle length 95-percent cumulative NDE
=
probability allowance for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC)
- Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.
(2)
The NDE is the value provided by the NRC in GL 95-05 as supplemented.
BEAVER VALLEY - UNIT 1 3/4 4-10d Amendment No.198
.DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.
The steam generator shall be determined OPERABLE after completing the corresponding a'ctions (plug or repair 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 each steam generator shall be reported to the Commission in a Special R4) ort pursuant to Specification 6.9.2.
b.
The complete results of the steam generator tube and sleeve 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 and sleeves inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged or repaired.
c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation.
The written report shall provide a description of investigations conducted to determine the 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 Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions 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.
2.
If circumferential crack-like indications are detected at the tube support plate intersections.
BEAVER VALLEY - UNIT 1 3/4 4-10e Amendment No.198
~ _ _
.DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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)
~2 voltage distribution exceeds 1 X 10 notify the Commission and provide an assessment of the safety significance of the occurrence.
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BEAVER VALLEY - UNIT 1 3/4 4-10f Amendment No.198 l
t L
o DPR-66 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Two Three Four Two Three Four I
First Inservice Inspection All One Two Two Second & Subsequent Inservice Inspections One (1)
One(1)
One (2)
One (3)
Table Notation:
(1)
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N percent of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other
[
steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
(2)
The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in (1) above.
(3)
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 i
inspections shall follow the instructions described in (1) above.
i BEAVER VALLEY - UNIT 1 3/4 4-10g Amendment No.jgg l
t DPR-66 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G.
C-2 Plug or repair C-1 None N/A N/A defective tubes and inspect additional C-2 Plug or repair defective C-1 None 2S tubes in this S.G.
tubes and inspect additional 4S tubes in C-2 Plug or repair this S.G.
defective tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 N/A N/A result of first sample C-3 Inspect all tubes in All other None N/A N/A this S.G.,
plug or S.G.s are repair defective C-1 tubes and inspect 2S tubes in each other S.G.
Some S.G.s Perform action for N/A N/A C-2 but no C-2 result of second Notification to NRC additional sample pursuant to S.G.s are Specification 6.6 C-3 Additional Inspect all tubes in S.G.
is each S.G.
and plug or N/A N/A C-3 repair defective tubes.
Notification to NRC pursuant to specifi-cation 6.6.
s=3Ht Where N is the number of steam generators in the unit, and n is the number of steam generators inspected n
during an inspection.
BEAVER VALLEY - UNIT 1 3/4 4-10h Amendment No.198
,DPR-66 4
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to:
a.
No pressure boundary LEAKAGE, b.
1 gpm unidentified LEAKAGE, c.
150 gallons per day primary-to-secondary LEAKAGE through any one steam generator, and d.
10 gpm identified LEAKAGE.
l APPLICABILITY:
MODES 1, 2,
3, and 4.
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 next 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, 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 />.
SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System LEAKAGES shall be denionstrated to be within each of the above limits by:
a.
Monitoring the following leakage detection instrumentation 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:(1) 1.
Containment atmosphere gaseous radioactivity monitor.
(1)
Only on leakage detection instrumentation required by LCO 3.4.6.1.
BEAVER VALLEY - UNIT 1 3/4 4-13 Amendment No.198 k
+
BASES 3/4.4.5 STEAM GENERATORS (Continued) 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 gallons per day per steam generator).
Axial cracks having a primary-to-secondary LEAKAGE less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants have demonstrated that primary-to-secondary LEAKAGE of 150 gallons per l
day per steam generator can readily be detected.
Leakage in excess j
of this limit will require plant shutdown and an unscheduled j
inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.
However, even if a defect of similar type should develop in service, it -will. be found during scheduled inservice steam generator tube examinations.
Plugging or I
repair will be required of all tubes with imperfections exceeding the plugging or repair limit.
Degraded steam generator tubes may be repaired by the installation of sleeves which span the degraded tube section.
A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not
- degraded, i
therefore, the sleeve is considered a part of the tube.
The surveillance requirements identify those sleeving methodologies approved for use.
If an installed sleeve is found to have through wall penetration greater than or equal to the plugging limit, the tube must be plugged.
The plugging limit for the sleeve is derived from R.G.
1.121 analysis which utilizes a 20 percent allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 2 09 percent of the original tube wall thickness.
The voltage-based repair limits of these surveillance requirements (SR) implement the guidance in Generic Letter (GL) 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections.
The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG.
Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no NDE detectable cracks extending outside the thickness of the support plate.
Refer to GL 95-05 for additional description of the degradation morphology.
Implementation of these SRs requires a derivation of the voltage structural limit from the burst versus voltage empirical BEAVER VALLEY - UNIT 1 B 3/4 4-2a Amendment No.198
.DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) 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 degradation growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit;
- VURL, is determined from the structural voltage limit by applying the following equation:
Vugt = VsL - Vor - Vgog where Vor represents the allowance for degradation growth between inspections and Vuos represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
The mid-cycle equation in SR 4.4.5.4.10.d should only be used during unplanned inspections in which addy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning the SGs to service.
For the 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 (EOC) voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service.
Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting
- criteria, then the results of the projected EOC voltage distribution should be provided per the GL section 6.b (c) criteria.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation.
Such cases will be considered by i
i 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.
BEAVER VALLEY - UNIT 1 B 3/4 4-2b Amendment No.
l l
1 i
g,DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAIUCE (Continued)
LCO (Continued) b.
Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period.
Violation of this LCO could result in continued degradation of the RCPB, if the LEARAGE is from the pressure boundary.
c.
Primarv-to-Secondary LEAKAGE throuah Any One SG Maintaining an operating LEAKAGE limit of 150 gpd per steam generator will minimize the potential for a large l
LEAKAGE event during a main steamline break.
Based on the non-destructive examination uncertainties, bobbin coil voltage distribution, and crack growth rate from the previous inspection, the expected leak rate following a steamline rupture is limited to below 4.5 gpm in the faulted loop.
Maintaining LEAKAGE within the 4.5 gpm limit will ensure that postulated offsite doses will remain within the 10 CFR 100 requirements and that control room habitability continues to meet GDC-19.
LEAKAGE in the intact loops will be limited to the operating limit of 150 gpd.
If the projected end-of-cycle distribution of crack indications results in primary-to-secondary LEAKAGE greater than 4.5 gpm in the l
faulted loop during a postulated steamline break event, additional tubes must be removed from service or repaired in order to reduce the postulated steamline break LEAKAGE to below 4.5 gpm.
l Also, the 150 gallons per day leakage limit incorporated into this specification is more restrictive than the standard operating leakage limit 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, it will be detected, and the plant shut down in a timely manner.
BEAVER VALLEY - UNIT 1 B 3/4 4-3g Amendment No.198
./ *DPR-66 REACTOR COOLANT SYSTEM
)
BASES
,3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) d.
Identified LEAKAGE l
Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not ir.terfere with detection of identified LEAKAGE and is well within the capability of the RCS Makeup System.
Identified LEAKAGE includes LEAKAGE to the containment i
from specifically known and located sources, but does not i
include pressure boundary LEAKAGE or controlled reactor l
coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).
Violation of this LCO could result in continued degradation of a component or system.
APPLICABILITY l
In MODES 1,
2, 3,
and 4,
the potential for RCPB LEAKAGE is greatest when the RCS ir pressurized.
In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses l
and reduced potentials for LEAKAGE.
LCO 3.4.6.3, "RCS Pressure Isolation Valve (PIV),* measures t
l leakage through each individual PIV and can impact this LCO.
Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight.
If both valves leak and result in a loss of mass from f
the RCS, the loss must be included in the allowable identified l
LEAKAGE.
l ACTIONS A.
If any pressure boundary LEAKAGE exists, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.
It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.
The allowed Completion Times are reasonable, based on i
operating experience, to reach the required plant conditions from full power conditions in an orderly manner and iithout challenging plant systems.
In MODE 5, the pressure tresses acting on the RCPB are much lower, and further deterioration is much less likely.
I BEAVER VALLEY - UNIT 1 B 3/4 4-3h Amendment No.*"
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