ML20198K735
| ML20198K735 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/21/1998 |
| From: | Capra R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198K739 | List: |
| References | |
| NUDOCS 9812310143 | |
| Download: ML20198K735 (14) | |
Text
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UNITED STATES ye 4
S NUCLEAR REGULATORY COMMISSION
- E If WASHINGTON, D.C. M1
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DUQUESK-LIGHT C.OMPANY OH') EDISON CO' APANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 219 License No. DPR-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duquesne Light Company, et al. (the licensee) dated June 18,1996, as supplemented September 8 and 30,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rufes 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.
l 9812310143 981221 PDR ADOCK 05000334 P
PDR l
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2-2.
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 Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 219, 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 loLfce..C u Robert A. Capra, Director Project Directorate 1-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 21, 1998 l
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l ATTACHMENT TO LICENSE AMENDMENT NO. 710 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 verticallines indicating the areas of change.
Remove Insert XVil XVil 3/4 4-10 3/4 4-10 3/4 4-10a 3/4 4-10a 3/4 4-10b 3/4 4-10b 3/4 410c 3/4 4-10c 3/4 4-10d 3/4 4-10d 3/4 4-10g 3/4 4-10g 3/4 4-1Ch 3/4 4-10h B 3/4 4-2b B 3/4 4-2b B 3/4 4-3f 8 3/4 4-3f 8 3/4 4-3g B 3/4 4-3g
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Table Index (cont.)
TABLE TITLE PAGE 3.3-11 Accident Monitoring Instrumentation 3/4 3-51 4.3-7 Accident Monitoring Instrumentation 3/4 3-52 Surveillance Requirements 3.3-13 Explosive Gas Monitoring Instrumentation 3/4 3-55 4.3-13 Explosive Gas Monitoring Instrumentation 3/4 3-57 Surveillance Requirements 4.4-1 Minimum Number of Steam Generators to be 3/4 4-10g l
Inspected During Inservice Inspection 4.4-2 Steam Generator Tube Inspection 3/4 4-10h l
4.4-3 Reactor Coolant System Pressure 3/4 4-14b Isolation Valves 3.4-1 Reactor Coolant System Chemistry Limits 3/4 4-16
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4.4-10 Reactor Coolant System Chemistry Limits 3/4 4-17 Surveillance Requirements 4.4-12 Primary Coolant Specific Activity Sample 3/4 4-20 and Analysis Program 3.7-1 Maximum Allowable Power Range Neutron Flux 3/4 7-2 High Setpoint With Inoperable Steam Line l
Safety Valves During 3 Loop Operation 3.7-2 Maximum Allowable Power Range Neutron Flux 3/4 7-3 l
High Setpoint with Inoperable Steam Line Safety Valves During 2 Loop operation 1
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3.7-3 Steam Line Safety Valves Per Loop 3/4 7-4 4.7-1 Snubber Visual Inspection Interval 3/4 7-31 4.7-2 Secondary Coolant System specific Activity 3/4 7-9 Sample and Analysis Program 3.8-1 Battery Surveillance Requirements 3/4 8-9a 3.9-1 Beaver Valley Fuel Assembly Minimum Burnup 3/4 9-15 vs. Initial U235 Enrichment For Storage in Region 2 Spent Fuel Racks I,
BEAVER VALLEY - UNIT 1 XVII Amendment No. 219
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%k SURVEILLANCE REQUIREMENTS (Continued)
Cateaorv Insoection Results C-1 Less than 5 percent 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 percent of the total tubes inspected are defective, or between 5
percent and 10 percent of the total tubes inspected are degraded tubes.
C-3 More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective.
Note:
In all inspections,. previously degraded tubes or sleeves must exhibit significant (greater than 10 percent) further wall penetrations to be included in the above percentage calculations.
The above required inservice 4.4.5.3 Insoection Frecuencies inspections of steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections following service under All Volatile Treatment (AVT) l conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
b.
If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose results 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 a subsequent inspection demonstrates that a third sample inspection is not required.
BEAVER VALLEY - UNIT 1 3/4 4-10 Amendment No. 219
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DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1.
Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, l
2.
A seismic occurrence greater than the Operating Basis Earthquake, 3.
A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.
A main steamline or feedwater line break.
l 4.4.5.4 Accentance Criteria a.
As used in this Specification:
1.
Imoerfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20 percent of the nominal tube wall thickness, if detectable, may be considered as imperfections.
j 2.
Dearadation means a
service-induced
- cracking,
- wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.
l 3.
Dearaded Tube means a
tube or sleeve containing.
j imperfections greater than or equal to 20 percent of 1
the nominal wall thickness caused by degradation.
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4.
Percent Dearadation means the percentage of the tube l
or sleeve wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it j
exceeds the p3ugging or repair limit.
A tube containing a defect is defective.
Any tube which does not permit the passage of the addy-current j
inspection probe shall be-deemed a defective tube.
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BEAVER VALLEY - UNIT 1 3/4 4-10a-Amendment No. 219 l
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=----m-;m-DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 6.
Pluccina or Renair Limit means the imperfection depth k
at or beyond which the tube shall be removed from r.wvice by plugging or repaired by sleeving in the a.tected 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%
l This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.
Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.
b)
ABB Combustion Engineering TIG welded l
sleeve wall 32%
c)
Westinghouse laser welded sleeve wall 25%
l 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 an Operating Basis Earthquake, a loss-of-coolant accident, or a steamline or feedwater line break as specified in 4.4.5.3.c, above.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot-leg side) completely around the U-bend to the top support to the cold-leg.
9.
Tube Reoair 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.
I The following sleeve designs have been found I
acceptable:
a)
ABB Combustion Engineering TIG Welded Sleeves, l
CEN-629-P, Revision 02 and CEN-629-P Addendum 1.
b)
Westinghouse laser welded sleeves, WCAP-13483, l
Revision 1.
s BEAVER VALLEY - UNIT 1 3/4 4-10b Amendment No. 219
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 10.
Tube Suonort 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 l
below:
a)
- tubes, whose degradation is l
attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or e<Iual to 2.0 volts will be allowed to remain in service.
b)
- tubes, whose degradation is l
attributed to outside diameter stress corrosion cracking within the bounds of the tube suppbrt 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 l
potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin l
voltage greater than 2.0 volts but less than or equal to the upper voltage repair limitW may remein 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 volta limit (ge) greater than the upper voltage repair will be plugged or repaired.
d)
If an unscheduled mid-cycle inspection is l
performed, the following mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.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. 219
i DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) a:===r The mid-cycle repair limits are determined from the following equations.
Vn j
Vmat =
' CL -- A t' 1.0 + NDE + Gr CL s s
Vuu = Vuvu-(Vat-Vat) CL-d '
(
CL s where:
upper voltage repair
- Vum,
=
limit YLRL lower voltage repair i
=
limit l
Vmna =
nid-cycle upper voltage repair limit based on i
time into cycle i
i Vxtnn =
nid-cycle lower voltage repair limit based on Vxuan and time into cycle
{
At length of time since
=
last scheduled inspection during which l
VuRL and V an were t
implemented cycle length (the time CL
=
between two scheduled steam generator inspections)
(
VsL structural limit voltage
=
Gr average growth rate per
=
cycle length l
NDE 95-percent cumulative
=
l probability allowance for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC)(2)
Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.10.a,
{
4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.
s (2)
The NDE is the value provided by the NRC in GL 95-05 as supplemented.
JEAVER VALLEY - UNIT 1 3/4 4-10d Amendment No. 219 A
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TABLE 4.4-1 DPR-66 l
MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE IN'iPECTION l
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1 freservice Inspection No Yes No. of Steam Generators per Unit
^ Three Three First Inservice Inspection All Two Second & Subsequent Inservice Inspections One (1)
One (2)
Table Notation:
(1)
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 9 percent of the tubes if the results of the first or previous inspections l
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.
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1 BEAVER VALLEY - UNIT 1 3/4 4-10g Amendment No. 219 1
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TABLE 4.4-2 DPR-66 o
L STEAM GENERATOR TUBE INSPECTION
[
b IST SANPLE INSPECTION 2ND SAMPLE INSPECTION 3R3 SANPLE INSPECTION Cample Size Result Action Required Result Action Required Result Action Required A minimum C-1 None N/A N/A N/A N/A (I
of C tubes C-2 Plug or repair defective C-1 None N/A N/A l
per S.G.
tubes and inspect C-2 Plug or repair defective C-1 None additional 25 tubes in tubes and inspect additional C-2 Plug or repair this S.G.
48 tubes in this S.G.
defective tubes C-3 Perform action
,a for C-3 result I;
of first sampla C-3 Perform action for C-3 result N/A N/A of first sample C-3 Inspect all tubes in this All other None N/A N/A S.G.,
plug or repair S.G.s are defective tubes and C-1 inspect 25 tubes in each Some S.G.s Perform action for C-2 result N/A N/A
'ther S.G.
C-2 but no of second sample o
additional Notification to NRC S.G.o are pursuant to Specification C-3 6.6 Additional Inspect all tubes in each N/A N/A S.G.
is S.G. and plug or repair C-3 defective tubes.
Notification to NRC pursuant to Specification 6.6.
e=9%
Where n is the number of steam generators inspected during an inspection.
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U BEAVER VALLEY - UNIT 1 3/4 4-10h Amendment No. 219' l
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i DPR-66 i '
BASES
]
I 2]MJ STEAM GENERATORS (Continued) 1 l
no NDE detectable cracks extending outside the thickness of the i
support plate.
Refar to GL 95-05 for additional description of the degradation morphology.
Implementation of these SRs requires a derivation of the voltage L
structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from j
the" structural limit (which is then implemented by this j
surveillance).
l The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95-percent prediction l
interval curve reduced to account for the lower 95/95-percent j
tolerance bound for tubing material properties at 650'F (i.e.,
the i
95-percent LTL curve).
The voltage structural limit must be adjusted j
downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty.
The upper i
voltage repair limit; Vm, is determined from the structural voltage j
limit by applying the following equation:
Vm = Vn - Vor - VNDE I
where Vcr represents the allowance for degradation growth between i
inspections and Vuos represents the allowance for potential sources i
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.
i Safety analyses were performed pursuant to Generic Letter 95-05 to i
determine the maximum MSLB-induced primary-to-secondary leak rate
{
that could occur without offsite doses exceeding a small fraction of j
10 CFR 100 (concurrent iodine spike), 10 CFR 100 (pre-accident iodine spike), and without control room doses exceeding GDC-19.
The current value of this allowable leak rate and a summary of the analyses are provided in Section 14.2.5 of the UFSAR.
The mid-cycle equation in SR 4.4.5.4.a.10.d should only be used l
during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommend'ed 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 l
BEAVER VALLEY - UNIT 1 B 3/4 4-2b Amendment No. 219
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t DPR-66 REACTOR COOLANT SYSTEM i
BASES f
3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) j APPLICABLE SAFFTY ANALYSES (Continued) affect the probability of such an event.
The safety analysis for an j
event resulting in steam discharge to the atmosphere conservatively assumes a 10 gpm primary-to-secondary LEAKAGE.
With the exception j
described below for the main steamline brea.k (MSLB) analyzed in l
support of voltage-based stcan generator tube repair criteria, j
Primary-to-secondary LEAKAGE is a
factor in the dose releases l
outside containment resulting from a steamline break (SL3) accident.
i To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube
- rupture (SGTR).
The leakage contaminates the secondary fluid.
The MSLB is more limiting for site radiation releases.
The primary-j to-secondary LEAKAGE assumed in the safety analysis for the MSLB' i
i accident is described in UFSAR Section 14.2.5.
The. radiological consequences of a MSLB outside of containment was reanalyzed in support of the tube support plate voltage-based repair criteria l
stated in SR 4.4.5.4.a.10.
For this analysis, the thyroid dose was maximized at 10% of the 10 CFR Part 100 guideline of 300 rem for the i
co-incident iodine spike case.
RCS leakage was based on projection i.
rather than on technical specification leakage limits.
The analysis j
indicated that offsite doses would remain within regulatory criteria i
with the assumed primary-to-secondary leakage (described in UFSAR j
Section 14.2.5) should steam generator tubes fail due. to the depressurization associated with a MSLB.
j A similar analysis was performed using a control room thyroid dose i
of 30 ::m as the criterion.
The control room was assumed to be j
manually isolated and pressurized at T=30 minutes for a period of one
- hour, at which time filtered emergency intake would be
. automat cally started.
The control room would be purged with fresh i
I air at T=8 hours following release cessation.
The analysis indicated that control room doses would remain within regulatory criteria with the assuned primary-to-secondary leakage (described in j
j UFSAR Section 14.2.5) should steam generator tubes fail due to the L
depressurization associated with a MSLB.
18.Q RCS operational LEAKAGE shall be limited to:
a.
Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.
LEAKAGE of this type is BEAVER VALLEY - UNIT 1 B 3/4.4-3f Amendment No. 219
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l, DPR-66 REACTOR COOLANT SYSTEM u
BASES 3/4.4.6.2 OPERATIONAL TF4XAGE (Continued) i LEO (Continued) l unacceptable as the leak itself could cause further i
deterioration, resulting in higher LEAKAGE.
Violation of i
this LCO could result in continued degradation of the RCPB.
LEAKAGE past seals and gaskets is not pres 3ure l
boundary LEAKAGE.
Should pressure boundary LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant
- System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation j
removes the source of potential failure.
1 l
b.
Unidentified T.EAKAGE 4
One gallon per minute (gpm) of unidentified LEAKAGE is i
allowed as a reasonable minimum detectable amount that the j
containment air monitoring and containment sump level monitoring equipment can detect within a reasonable t,ine i
period.
Violation of this LCO could result in continued l
degradation of the RCPB, if the LEAKAGE is from the j
l pressure boundary.
j c.
Primarv-to-Secondary LEAKAGE throuah Any One SG l
Operating experience at PWR plants has shown that sudden increases in leak rate are often precursors to larger tube failures.
Maintaining an operating LEAKAGE limit of 150 gpd per steam generator will minimize the potential for a large LEAKAGE event at power.
This operating LEAKAGE limit is more restrictive than the operating LEAKAGE limit in st'andardized technical specifications.
This provides additional margin to accommodate a tube flaw which might grow at a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate.
This reduced LEAKAGE limit, in conjunction with a leak rate monitoring program, provides additional assurance that this precursor LEAKAGE will be detected and the plant shut down in a timely manner.
1 j
BEAVER VALLEY - UNIT 1 B 3/4 4-3g Amendment No. 219 I
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