ML20138J751
| ML20138J751 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/05/1997 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20138J755 | List: |
| References | |
| NUDOCS 9705080327 | |
| Download: ML20138J751 (38) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20066 4001
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DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE t
Amendment No.
175 License No. NPF-9 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-9 filed by the Duke Power Company (licensee) dated September 30, 1994, as supplemented by letters dated September 18, 1995, and March 15, April 29, May 16, September 23, and October 28, 1996, and January 16, April 22, and May 2,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in i
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 set forth in 10 CFR Chapter I; 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.
1 9705000327 970505 PDR ADOCK 05000369 P
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- l 2.
Accordingly, the license is hereby amended by page changes to the i
Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) lows:of Facility Operating License No. NPF-9 i
j is hereby amended to read as fol l
Technical Soecifications
)
l The Technical Specifications contained in Appendix A, as revised j
through Amendment No.175
, are hereby incorporated into this I
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 and shall be implemented within 30 days from the date of issuance.
F THE NUCLEAR REGULATORY COMISSION i
I6k.
l Her ert N. Berkow, Director i
Project Directorate II-2 Division of Reactor Projects - I/II l
Office of Nuclear Reactor Regulation 4
i i
Attachment:
i Technical Specificat! ion Changes Date of Issuanca:
May 5, 1997
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S
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ATTACHMENT TO LICENSE AMENDMENT NO.
175 i
FACILITY OPERATING LICENSE NO. NPF-9 i
DOCKET NO. 50-369 l
I Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised'pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Paaes Insert Paaes 2-5 2-5 2-8 2-8 2-10 2-10 3/4 3-29 3/4'3-29 3/4 3-30 3/4 3-30 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14 4
3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-20 3/4 4-20 B 3/4 4-3 8 3/4 4-3 i
i B 3/4 4-4 B 3/4 4-4 8 3/4 4-5 B 3/4 4-5 5-7 5-7 6-22 6-22 4
l I
j i
a i
i
TABLE 2.2-I (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 12. Steam Generator Water 216.7% of span 2 15% of span l
Level--Low-Low
- 13. Undervoltage-Reactor 2 5082 volts-each bus 1 5016 volts-each bus Coolant Pumps
- 14. Underfrequency-Reactor 2 56.4 Hz - each bus 2 55.9 Hz - each bus Coolant Pumps
- 15. Turbine Trip a.
Low Trip System Pressure 2 45 psig 2 42 psig b.
Turbine Stop Valve Closure 2 1% open 2 1% open
- 16. Safety Injection Input N.A.
N.A.
from ESF
- 17. Reactor Trip System Interlocks a.
Intermediate Range Neutron Flux, P-6, 2 1 x 10'" amps 2 6 x 10'" amps Enable Block Source Range Reactor Trip McGUIRE - UNIT 1 2-5 Amendment No.175
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TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
-.i NOTATION (Continued)
NOTE 1:
(Continued) t, Time constant utilized in the measured T, lag compensator, as preser.ted in the Core Operating Limits Report, T'
s 585.l*F Reference T, at RATED THERMAL POWER, l
l Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in 4
=
the Core Operating Limits Report, P
Pressurizer pressure, psig, P'
2235 psig (Nominal RCS operating pressure),
Laplace transform operator, sec",
f s
and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range J
t nuclear ton chambers; with gains to be selected based on measured instrument response during plant startup
[
tests such that:
(1) for q, - q, between the " positive" and " negative" f (AI) breakpoints as presented in the Core j
t Operating Limits Report; f (AI) = 0, where q, and q, are percent RATED THERMAL POWER in the top and 1
bottom halves of the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL t
POWER; t
(11) for each percent imbalance that the magnitude of q, - g is more negative than the f (AI) " negative" t
breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f (AI) " negative" slope presented in the Core Operating Limits Report; l
1 and (iii) for each percent imbalance that the magnitude of q, - g is more positive than the f (AI) " positive" t
breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f (AI) " positive" slope presented in the Core Operating Limits Report.
t i
l McGUIRE - UNIT I 2-8 Amendment No. 175 r
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TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
As defined in Note 1, T
s 585.1*F Reference T,at RATED THERMAL POWER, l
T" As defined in Note 1, and s
f (AI) is a function of the indicated difference between top and bottom detectors of the power-range 2
nuclear ton cha=bers; with gains to be selected based on measured instrument response during plant startup tests such that:
(1) for q, - q, between the " positive" and ' negative" f (AI) breakpoints as presented in the Core 2
Operating Limits Report; f (AI) - 0, where q, and q, are percent RATED THERMAL POWER in the 2
top and bottom halves of the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of q, - q, is more negative than the f (AI) 2
" negative" breakpoint presented in the Core Operating Limits Report, the AT Trip.Setpoint shall be automatically reduced by the f (AI) " negative" slope presented in the Core Operating Limits Report; and (iii) for each percent imbalance that the magnitude of q, - q, is more positive than the f (AI) 2
" positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint i
shall be automatically reduced by the f (AI) " positive" slope presented in the Core Operating 2
Limits Report.
i NOTE 3:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.4% of Rated Thermal Power.
NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of Rated Thermal Power.
McGUIRE - UNIT 1 2-10 Amendment No. 175
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TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS fjNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 4.
Steam Line Isolation a.
Nanual Initiation M.A.
N.A.
b.
Automatic Actuation Logic N.A.
N.A..
and Actuation Relays c.
Containment Pressure--High-High s 2.9 psig s 3.0 psig
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d.
Negative Steam Line s 100 psi with a s 120 psi with a Pressure Rate - High rate / lag function rate / lag function 3
time constant time constant 2 50 seconds 2 50 seconds i
e.
Steam Line Pressure - Low 2 775 psig 2 755 psig 5.
Turbine Trip and Feedwater Isolation a.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays b.
Steam Generator Water level--
s 83.9% of narrow range s 85.6% of narrow range l
High-High (P-14) instrument span each steam instrument span each steam generator generator c.
Doghouse Water Level-High 12" 13" (Feedwater Isolation Only) 6.
Containment Pressure Control System Start Pemissive/ Termination 0.3 $ SP/T s 0.4 PSIG 0.25 s SP/T s 0.45 PSIG (SP/T)
McGUIRE - UNIT 1 3/4 3-29 Amendment No. 175
=.
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 7.
Manual Initiation N.A.
N.A.
b.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays c.
Steam Generator Water Level--Low-Low 1)
Start Motor-Driven Pumps 216.7% of span 215% of span 2)
Start Turbine-Driven Pumps 216.7% of span 2 15% of span d.
Auxiliary Feedwater 2 2 psig 2 1 psig Suction Pressure - Low (Suction Supply Automatic Realignment) e.
Safety Injection -
See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps and Allowable Values McGUIRE - UNIT 1 3/4 3-30 Amendment No.17S
= _ - -.
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REACTOR COOLANT S.1111ti 4
SURVEILLANCE RE0VIREMENTS (Continued) a i
1)
All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
j 2)
Tubes in those areas where experience has indicated potential problems, and 4
i 3)
A tube inspection (pursuant to Specification.4.4.5.4.a.8) shall be performed on each selected tube.
If any selected tube does l
not permit the passage of.the eddy current probe for a tube J
inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
s The tubes selected as the second and third samples (if required by c.
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 where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Cateaory Insoection 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 I
significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
McGUIRE - UNIT 1 3/4 4-13 Amendment No.175
e REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.3 Insnection Freauencies - The above required inservice ins'pections of steam generator tubes shall be performed at the following frequencies:
The first inservice inspection after the steam generator replacement I
a.
shall be performed after at least 6 Effective Full Power Months but I
within 24 calendar months of initial criticality after the steam generator replacement. Subsequent inservice inspections shall be 4
performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive i
inspections following service under AVT 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 results of the inservice inspection of a steam generator s
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 extended to a maximum of once per 40 months; and Additional, unscheduled inservice inspections shall be performed on c.
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 tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, 2)
A seismic occurrence greater than the Operating Basis Earthquake, 3)
A loss-of-coolant accident requiring actuation of the Engineered Safety Features, and 4)
A main steam line or feedwater line break.
McGUIRE - UNIT 1 3/4 4-14 Amendment No.175
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e s
RFACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 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 from that required by fabrication drawings or l
specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be consid-l ered as imperfections; 2)
Dearadation means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube; I
3)
Dearaded Tubt means a tube containing imperfections greater than or equal to 20% of the nominal tube wall thickness caused by degradation; 4)
% dearadation means the percentage of the tube wall thickness l
affected or removed by degradation; j
5)
Defect means an imperfection of such severity tilat it exceeds the plugging limit. A tube containing a defect is defective; l
6)
Pluaaina Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging and is equal to 40% of the nominal tube wall thickness.
7)
Unserviceable describes the condition of a tube if it leaks or i
contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above; 8)
Tube Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and McGUIRE - UNIT 1 3/4 4-15 Amendment No.175
_ _ =
1 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) i 9)
Preservice Insoection means an inspection of the full length of each tube in each steam generator performed by eddy current i
techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
1 b.
The steam generator shall be determined OPERABLE after completing
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the corresponding actions (plug all tubes exceeding the plugging l
limit and all tubes containing through-wall cracks) required by j
Table 4.4-2.
4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged in each steam j
generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1)
Number and extent of tubes inspected, 2)
Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged.
l 3
t' McGUIRE - UNIT 1 3/4 4-16 Amendment No.175
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 First Inservice Inspection After Steam Generator All One Two Two Replacement Second & Subsequent Inservice Inspections One' One One One 2
3 TABLE NOTATION:
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where H 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 after steam I
i generator replacement shall be inspected. The third and subsequent inspections should follow the I
instructions described in I above.
3 Each of the other two steam generators not inspected during the first inservice inspections after steam generator replacement shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in I above.
McGUIRE - UNIT 1 3/4 4-17 Amendment No. 175 i
j REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE l
LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
I gpm UNIDENTIFIED LEAKA'GE, c.
0.27 gpm total primary-to-secondary leakage through all steam generators and 135 gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.
2235 1 20 psig, and f.
I gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a.
With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> arei in COLD SHUTDOWN within the following i
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 /> Py use of at least two closed manual 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 />.
McGUIRE - UNIT 1 3/4 4-20 Amendment No. 175
BASES 3/4.4.4 RELIEF VALVES (Continued) reactor coolant pressure boundary leakage. 3) Manual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of RCS pressure and isolate a PORV with excessive leakage.
- 4) Automatic control of PORVs to control RCS pressure. This is a function that reduces challenges to the code safety valves for overpressurization events.
- 5) Manual control of a block valve to isolate a stuck-open PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for -inservice 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 surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage - 135 gallons per day per steam generator). Cracks having l
a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 135 gallons per day per steam generator can l
readily be detected.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it McGUIRE - UNIT I B 3/4 4-3 Amendment No. 175
l l
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) i I
will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably d
detect wastage type degradation that has penetrated 20% of the original tube wall thickness.
a Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-
)
sion pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of plant 3
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, i
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 provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of i
i Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced
?
to a threshold value of less than I gpm. This threshold value is sufficiently j
low to ensure early detection of additional leakage.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross y
l 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.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited i
amount of leakage from known sources whose presence will not interfere with i
the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
i i
i i
j McGUIRE - UNIT I B 3/4 4-4 Amendment No. 175
BASES OPERATIONAL LEAKAGE (continued)
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.
i This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.
)
The total steam generator tube leakage limit of 0.27 gpm for all steam 4
i generators and the 135 gpd leakage limit per generator ensures that the dosage i
contribution from the tube leakage will be limited to the applicable fraction of 10 CFR Part 100 dose guideline values for all FSAR Chapter 15 transients.
l The 0.27 gpa and the 135 gpd limits are consistent with the assumptions used in the analysis of these accidents. The 135 gpd leakage limit per steam generator i
4 ensures that steam generator tube integrity is maintained in the event of a 2
l main steam line rupture or under LOCA conditions.
l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. There-i fore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be
{
promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEMISTRY d
The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-i tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the. Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that l
o>eration may be continued with contaminant concentration levels in excess of j
tie Steady State Limits, up to the Transient Limits, for the specified limited l
time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued opera-i tion within the restrictions of the Transient Limits provides time for taking i
corrective actions to restore the contaminant concentrations to within the l
Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take j
corrective ACTION.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure 3
i that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 0.27 gpm.
The values for l
the limits on specific activity represent limits based upon a parametric McGUIRE - UNIT 1 B 3/4 4-5 Amendment No. 175
DESIGN FEATURES FUEL ASSEMBLIES Continued)
J to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL R0D ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 i
inches of absorber material. The nominal values of absorber material for Unit I control rods shall be 80% silver,15% indium, and 5% cadmium. All control rods shall be clad with stainless steel tubing.
I 5.4 REACTOR COOLANT SYSTEM e
DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
t a.
In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, j
b.
For a pressure of 2485 psig, and c.
For a temperature of 650'F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coo 1&nt System is 13,050 l i 100 cubic feet at a nominal T,, of 525'F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 a.
The spent fuel storage racks are designed and shall be maintained with:
1) k s 0.95 if fully flooded with unborated water as d,e,s,cribed in Section 9.1 of the FSAR; and 2)
A nominal 10.4" center to center distance between fuel assemblies placed in Region 1; and 3)
A nominal 9.125" center to center distance between fuel assemblies placed in Region 2.
McGUIRE - UNIT 1 5-7 Amendment No.175
i
~ ADMINISTRATIVE CON 1ROLS f
CORE OPERATING LIMITS REPORT (Continued) 8.
DPC-NE-3002, Rev 1, "FSAR Chapter 15 System Transient Analysis Methodology," SER dated December 1995.
t (Methodology used in the system thermal-hydraulic analyses which determine the core operating limits) j 9.
DPC-NE-3000P, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology,"
i SER dated December 1995.
(Modeling used in the system thermal-hydraulic analyses) 10.
DPC-NE-1004A, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
November, 1992.
l (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
11.
DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC j
Proprietary).
]
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear i.
Enthalpy Rise Hot Channel Factor FoH(X,Y).)
12.
DPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
13.
DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,".
February 1995 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor).
14.
BAW-10162P-A, TACO 3 Fuel Pin Thermal Analysis Computer Code, B&W Fuel Company, November 1989.
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
15.
BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.
(Used for Specification 2.2.1, Reactor Trip System Instrumentation i
Setpoints).
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
McGUIRE - UNIT 1 6-22 Amendment No. 175
Y *s3o9 k
UNITED STATES g
j NUCLEAR REGULATORY COMMISSION l
2 WASHINGTON, D.C. enmas nnnt 49.....
DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 157 License No. NPF-17 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated September 30, 1994, as supplemented by letters dated September 18, 1995, and March 15, April 29, May 16, September 23, and October 28, 1996, and i
January 16, April 22, and May 2, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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; j
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.
i e
. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.
NPF-17 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.157
, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment shall be effective upon replacement of the steam generators.
FOR THE NUCLEAR REGULATORY COMMISSION LA H rbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
May 5,1997
}
c 4
h 4
4
ATTACHMENT TO LICENSE AMENDMENT NG.157 FACILITY OPERATING LICENSE NO. NPF-17 QQCKET NO. 50-370 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified b contain vertical lines indicating the areas of change. y Amendment number and Remove Paaes Insert Paaes 2-5 2-5 2-8 2-6 2-10 2-10 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-20 3/4 4-20 B 3/4 4-3 B 3/4 4-3 B 3/4 4-4 B 3/4 4-4 8 3/4 4-5 B 3/4 4-5 5-7 5-7 6-22 6-22
~.
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 12. Steam Generator Water 216.7% of span 215% of span l
Level--Low-Low
- 13. Undervoltage-Reactor 2 5082 volts-each bus 2 5016 volts-each bus Coolant Pumps
- 14. Underfrequency-Reactor 2 56.4 Hz - each bus 2 55.9 Hz - each bus Coolant Pumps
- 15. Turbine Trip a.
Low Trip System Pressure 2 45 psig 2 42 psig b.
Turbine Stop Valve Closure 2 1% open 2 1% open
- 16. Safety Injection Input N.A.
N.A.
from ESF
- 17. Reactor Trip System Interlocks a.
Intermediate Range Neutron Flux, P-6, 2 1 x 10'" amps 2 6 x 10'" amps Enable Block Source Range Reactor Trip McGUIRE - UNIT 2 2-5 Amendment No. 157 I
-. ~.
~.
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
NOTE 1:
(Continued)
Time constant ctilized in the measured T lag compensator, as presented in the Core Operating x,
m Limits Report, s 585.1*F Reference T,at RATED THERMAL POWER, l
T' Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in K
3 the Core Operating Limits Report, Pressurizer pressure, psig, P
P' 2235 psig (Nominal RCS operating pressure),
Laplace transform operator, sec",
s and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range 1
nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup s
tests such that:
(1) for q, - g between the " positive" and " negative" f (AI) breakpoints as presented in the Core 1
Operating Limits Report; f (AI) - 0, where q,.and q, are percent RATED THERMAL POWER in the top and t
bottom halva of the core respectively, and q, + g is total THERMAL POWER in percent of RATED THERMAL POWER; (11) for each percent imbalance that the magnitude of q, - g is more negative than the f (AI) " negative" t
breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be i
automatically reduced by the f (AI) " negative" slope presented in the Core Operating Limits Report;
[
t and
[
I (iii) for each percent imbalance that the magnitude of q, - g is more positive than the f (AI) " positive" 1
breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f (AI) " positive" slope presented in the Core Operating Limits Report.
t McGUIRE - UNIT 2 2-8 Amendment No.157 i
i
.- l TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
As defined in Note 1 T
{
T' s 585.l*F Reference T,at RATED THERMAL POWER, l
=
As defined in Note 1, and s
i f (AI) is a function of the indicated difference between top and bottom detectors of the power-range 2
nuclear ton chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(1) for q, - g between the " positive" and " negative" f (AI) breakpoints as presented in the Core 2
Operating Limits Report; f (AI) = 0, where q, and g are percent RATED THERMAL POWER in the 2
top and bottom halves of the core respectively, and q, + g is total THERMAL POWER in percent of l
RATED THERMAL POWER-t (11) for each percent imbalance that the magnitude of q, - g is more negative than the f (AI) 2
" negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint l
shall be automatically reduced by the f (AI) " negative" slope presented in the Core Operating 2
Limits Report; and t
(;11) for each percent imbalance that the magnitude of q, - g is more positive than the f (AI) 2
" positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f (AI) " positive" slope presented in the Core Operating 2
Limits Report.
i NOTE 3:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.4% of Rated Thermal Power.
NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of Rated t
Thermal Power.
l f
r McGUIRE - UNIT 2 2-10 Amendment No. 157 l
r t
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 4.
Steam Line Isolation a.
Manual Initiation N.A.
N.A.
b.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays c.
Containment Pressure--High-High s 2.9 psig s 3.0 psig d.
Negative Steam Line s 100 psi with a s 120 psi with a Pressure Rate - High rate / lag function rate / lag function time constant time constant t 50 seconds 1 50 seconds e.
Steam Line Pressure - Low 2 775 psig 2 755 psig 5.
Turbine Trip and Feedwater Isolation a.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays b.
Steam Generator Water level--
s 83.9% of narrow range s 85.6% of narrow range l
High-High (P-14) instrument span each steam instrument span each steam generator generator c.
Doghouse Water Level-High 12" 13" (Feedwater Isolation Only) j 6.
Containment Pressure Control System Start Permissive / Termination 0.3 $ SP/T s 0.4 PSIG 0.25 s SP/T s 0.45 PSIG (SP/T)
McGUIRE - UNIT 2 3/4 3-29 Amendment No. 157
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 7.
Manual Initiation N.A.
N.A.
b.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays c.
Steam Generator Water Level--Low-Low 1)
Start Motor-Driven Pumps 216.7% of span 2 15% of span 2)
Start Turbine-Driven Pumps 2 16.7% of span 215% of span d.
Auxiliary Feedwater 2 2 psig 2 1 psig Suction Pressure - Low (Suction Supply Automatic Realignment) e.
Safety Injection -
See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps and Allowable Values McGUIRE - UNIT 2 3/4 3-30 Amendment No.
157
SURVEILLANCE REQUIREMENTS (Continued) i 1)
All nonplugged tubes that previously had detectable wall j
penetrations (greater than 20%),
~
i i
2)
Tubes in those areas where experience has indicated potential problems, and i
J 3)
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube 3
inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
c.
The tubes selected as the second and third samples (if required by i
Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
j; 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.
The results of each sample inspection shall be classified into one of the following three categories:
Cateaory Insoection 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 j
total tubes inspected are defective, or between 5% and 10% of the total. tubes inspected are degraded tubes.
C-3 Mor,e 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.
l McGUIRE - UNIT 2 3/4 4-13 Amendment No. 157
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
I 4.4.5.3 Insoection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
The first inservice inspection after the steam generator replacement a.
shall be performed after at least 6 Effective Full Power Months but within 24 calendar months of initial criticality after the steam generator replacement. 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 AVT 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 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 extended to a maximum of once per 40 months; and c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection 2
specified'in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1)
Reactor-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the
. limits of Specification 3.4.6.2, 2)
A seismic occurrence greater than the Operating Basis Earthquake, 3)
A loss-of-coolant accident requiring actuation of the Engineered Safety Features, and 4)
A main steam line or feedwater line break.
4 J
b 4
McGUIRE - UNIT 2 3/4 4-14 Amendment No.157
}
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
.i 4.4.5.4 Accentance Criteria a.
As used in this specification:
1 1)
Innerfection means an exception to the dimensions, finish or j
contour of a tube from that required by fabrication drawings or l
specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be consid-l l
ered as imperfections; 2)
Dearadation means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube; l
l 3)
Dearaded Tube means a tube containing imperfections greater l
than or equal to 20% of the nominal tube wall thickness caused by degradation;
).
4)
% dearadation means the percentage of the tube wall thickness l
affected or removed by degradation; l
5)
Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; l
t 6)
Pluaaina Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging and is equal to 40% of the nominal tube wall thickness.
4 7)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as 1
specified in 4.4.5.3c, above; i
8)
Tube Insoection means an inspection of the steam generator tube i
from the point of entry (hot leg side) completely around the U-i bend to the top support of the cold leg; and i
4 j
McGUIRE - UNIT 2 3/4 4-15 Amendment No.157
. - ~,,
)
SURVEILLANCE REQUIREMENTS (Continued)
I 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition 4
of the tubing. This inspection shall be performed prior to I
initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
4 b.
The steam generator shall be determined OPERABLE after completing i
the corresponding actions (plug all tubes exceeding the plugging l
limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reoorts Within 15 days following the completion of each inservice inspection a.
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; I
b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant % Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1)
Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and i
j 3)
Identification of tubes plugged.
l 4
t 1
j i
McGUIRE - UNIT 2 3/4 4-16 Amendment No. 157
TABLE 4.4-1 MININUM 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 First Inservice Inspection After Steam Generator All One Two Two Replacement Second & Subsequent' Inservice Inspections One' One' One One z
3 TABLE NOTATION:
ine inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 T, % 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 after steam generator replacement 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 after steam generator replacement shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.
McGUIRE - UNIT 2 3/4 4-17 Amendment No. 157
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
I gpm UNIDENTIFIED LEAKAGE, c.
0.27 gpm total primary-to-secondary leakage through all steam generators and 135 gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.
2235 i 20 psig, and f.
I gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
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 following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With any Reactor Coolant System Pressure Isolation Valve leakage c.
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 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 />.
l McGUIRE - UNIT 2 3/4 4-20 Amendment No. 157 1
. ~ _ _ _ _ _ _ -_
BASES i
3/4.4.4 RELIEF VALVES (Continued) reactor coolant pressure boundary leakage. 3) Nanual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of RCS pressure and isolate a PORV with excessive leakage. 4) Automatic control of PORVs to control RCS pressure.
This is a funct e that reduces challenges to the code safety valves for overpressurization events,
- 5) Manual control of i
a block valve to isolate a stuck-open PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice 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 surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage - 135 gallons per day per steam generator). Cracks having l
a reactor-to-secondary leakage less than this limit during operation will have
'an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 135 gallons per day per staam generator can l
readily be detected.
Leakage in excess of this limit wili require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with pro >er chemistry treatment of the secondary coolant. However, even if a defect siould develop in service, it 1
l l
l McGUIRE - UNIT 2 B 3/4 4-3 Amendment No. 157 I
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) will be fotnd during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type 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 promptly reported to the Commis-sion pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of plant operation.
Such cases will be considered by the Commission on a case-by-case I
basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revisipn of the Technical i
Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS l
The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coo 1 M. pressure l
boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection i
Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE l
l Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than I gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
i The Surveillance Requirements for RCS pressure isolation valves provide 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 t,e considered as a portion of the allowed limit.
The 10 gpa 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.
i 3
l l
McGUIRE - UNIT 2 B 3/4 4-4 Amendment No.157 1
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (continued)
The CONTROLLED LEAKAGE 'iimitation restricts operation when the total flow supplied to the reactor coolait pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
l This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.
The total steam generator tube leakage limit of 0.27 gpm for all steam generators and the 135 gpd leakage limit per generator ensures that the dosage contribution from the tube leakage will be limited to the applicable fraction of 10 CFR Part 100 dose guideline values for all FSAR Chapter 15 transients.
The 0.27 gpm and the 135 gpd limits are consistent with the assumptions used in the analysis of these accidents. The 135 gpd leekage limit per steam generator ensures that steam generator tube integrity is aaintained in.the event of a main steam line rupture or under LOCA conditions.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failura of the pressure boundary. There-fore, the presence of any PRESSURE BOUND'4RY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEMISTRY l
l The limitations on Reactor Coolant System chemistry ensure that corrosion l
of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life t
of the plant. The associated effects of exceeding the oxygen, chloride, and l
fluoride limits are time and temperature dependent. Corrosion studies show that i
operation may be continued with contarsinant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued opera-tion within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take corrective ACTION.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 doce guideline values following a steam j
generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 0.27 gpm.
The values for l
j the limits on specific activity represent limits based upon a parametric McGUIRE - UNIT 2 B 3/4 4-S Amendment No. 157 I
w
.-y
i DESIGN FEATURES FUEL ASSEMBLIES Continued) to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
l l
CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material for Unit 2 control rods shall be 100% boron carbide (B 0) for 102 inches and 80%
4 silver, 15% indium, and 5% cadmium for the 40-inch tip. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGNPRESSUREdNDTEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
a.
In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 650'F, except for the pressurizer which is 680'F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,050 l 100 cubic feet at a nominal T, of 525'F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 a.
The spent fuel storage racks are designed and shall be maintained with:
1) k s 0.95 if fully flooded with unborated water as d,e,s,cribed in Section 9.1 of the FSAR; and 2)
A nominal 10.4" center to center distance between fuel i
assemblies placed in Region 1; and 3)
A nominal 9.125" center to center distance between fuel l
assemblies placed in Region 2.
4 McGUIRE - UNIT 2 5-7 Amendment No. 157 i
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) l 8.
DPC-ME-3002, Rev 1, "FSAR Chapter 15 System Transient Analysis Methodology," SER dated December 1995.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits) 9.
DPC-NE-3000P, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology,"
l SER dated December 1995.
(Modeling used in the system thermal-hydraulic analyses) 10.
DPC-NE-1004A, " Nuclear Design Methodology Using CASM0-3/ SIMULATE-3P,"
l November, 1992.
(Methodology for Specification 3.1.1.3 - Moderator Temperature j
Coefficient.)
11.
DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC l
Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear l
Enthalpy Rise Hot Channel Factor FoH(X,Y).)
12.
DPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation l
Setpoints.)
j 13.
DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"
February 1995 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reacter Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor).
14.
BAW-10162P-A, TAC 03 Fuel Pin Thermal Analysis Computer Code, B&W Fuel Company, November 1989.
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
15.
BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.
i l
(Used for Specification 2.2.1, Reactor Trip System Instrumentation l
Setpoints).
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
McGUIRE - UNIT 2 6-22 Amendment No. 157 i