ML20012E782
| ML20012E782 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/28/1990 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20012E783 | List: |
| References | |
| NUDOCS 9004060294 | |
| Download: ML20012E782 (16) | |
Text
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UNITED sT AT E s
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NUCLE AR REGULATORY COMMISSION WALHINGT ON, D. C. 70146
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DV_KE POWER Colf ANY NORTH Ct.ROLINA ELECTRIC MEMBERSHIP CORPORATION
$ ALUDA RIVER ELECTRIC COOPERATIVE,1h0.
DOCIET N0. 50-413 CAT AWB A NUCLEAR STATION, UNIT 1 AMENCl'ENT TO F ACILITY OPERATING LICENSE Amendrent No. 73 License No. Ftf-35 1.
The Nuclear Regulatory Comission (the Connission) has found thatt A.
The application for amendment to the Cataba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Ccepany acting for itself, North Carolina Electric Mertiership) Corporation and Saluda River Electric Cooperative, Inc.,
(licensee s dated April 19, 1989, as supplerented Septeiter 13, 1989, complies with the standards and re Energy Act of 1954, as anended (the Act)quirenents of the Atomic, and the Comissie and regulations as set forth in 10 CTR Chapter It 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 (i) that the activities authorized by thit amendnent 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 set forth in 10 CFR Chapter I; D.
The issuance of this arenament will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this atendment is in accordance with 10 CFR Part 51 of the Cornission's regulations and all applicable requirements have been satisfied, i
9004060294 9003:2 FDR ADOCK 0".000413 P
i 2
E.
Accordirgly, the lictnse is hereby amended by page changes to the Technical
$recifications as indicated in the attachrrent to this liter.se amend.ent, and Paragraph 2.C.(2) of Tacility Operating License 140. NPT.S$ is hereby amended to read as follows.
Technical Specificationc The Technical Specifications contained in /sppendix A, as revised threugh Anendnent Io 73
, are hereby incorporated into the license.
The licensee shall operate tie facility in accordance with the Technical Specifications and the Environrental Protection Plan.
3.
This license amendnent is effective cs of its date of issuance.
FCt DIE 10 CLEAR REGULATORY C0tti!SSION lY' W
s s Dav id C. I'.a tt hew s, Di re ctor Project Directorate 113 Division of P.eactor Projects.1/11 Office of Nuclear Reactor R(gulatien Attac hnent:
Technic 61 Specification Changes Date of Issuance:
P, arch 28, 1990
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- e UNITtD STATES l'
i NUCLE AR RE^UL ATORY COMMISSION l
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utmwo f ow, o. c. rot.n i
(,..... f DL'KE POWER C0ff At?Y NmTH CAROLINA MJNICIPAL POWER ACDiCY NO.1 i
PIEDMONT ML'NICIPAL POWER AGENCY f
DOCKET NO. $0 414 CATAWBA NUELEAR STATICN. LINIT 2 I
AMEtLMENT TO FACILITY OPEPATING LICEN$E j
i Amendaent No. 67
}
License No. NFF.52 i
1.
The Nuclear Pegulatory Commission (the Commission) has found that:
A.
The application for amendnent to the Catawba Nuclear Station, Unit 2 I
(the facility) facility Operating License No NPF 5? filed by the Du6e Power Company acting for itself, North Carolina Municipal Power i
Agency No. I and Piedmont Municipal Power Agency, (licensees) dated April 19,1989, as supplemented Septeiser 13, 1989, coglies 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; l
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the j
Copenission; t
C.
There is reasonable assurance (1) that the activities authorized by this amendnent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cogliar.ce with the Consnission's regulations set forth l
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 amendnent is in accordance with 10 CFR Part $1 of the Commission's regulations and all applicable requirenents have been satisfied.
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Accordingly, the license is hereby anended by page changes to the Technical i
$pecifications as indicated in the attachment to this license amendment.
and Paragraph 2.C.(2) of facility Operating License No. NPT.52 is herely an, ended to read as follows:
i Technical Specifications i
The Technical Specifications contained in Appendix A, as revised through Anendrent No. 67, are hereby incorporated into the license.
The licensee shall cperate the facility in accordance with the i
Technical Specifications and the Environnental Pmtection Flan.
3.
This license anendnent is effective as of its date of issuance, j
TOR THE NUCLEAR REGULATORY COMMIS$10N i
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Cavid B. Matthews, Director l
Project Directorate !!.3 Division of Reactor Projects - 1/l!
Office of Nuclear Reactor Regulation Attachnent:
l Technical Specification Changes 1
Date of issuance: March 28, 1990 f
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ATTAC *ENT TO L! CENSE AMEN!HENT NO. 73 i
TACILITY CFERATING LICENSE NO. Nrf 35 I
DOCKET NO. 50 413 i.
AND TO LICENSE AMENDMENT N0. 67 L
FACILITY OPERATit;G LICENSE t:0. NPr 50 l
D0CKET NO. 50-414 l
Replace the following pages of the Appendix "A" Technical Specifications with i
the enclosed pages.
The revittd pages are identified by Amendment riup6er and i
contain vertical lir.es indicating the areas of change.
The corresponding wer.
1er.f pages are also provided to mair.tain document coupleteness, i
i Anwneed Page Overleaf Page l
XIV XI!!
3/4 4-32 3/4431 3/4 4 33 i
3/4 4-34 i
3/4435 i
B3/4 4 8 03/4 4 7 B3/4 412 B3/4 4-11 l
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y BASES t
SECTION PAGE t
[
7 L
TABLE B 3/4.4 1 REACTOR VESSEL TOUGHNESS.........................
B 3/4 4-9 l
p' FIGURE B 3/4.4 1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF l
FULL POWER SERVICE LIFE..................................
B 3/4 4 11 g
3/4.4.10 STRUCTURAL INTEGRITY.....................................
B 3/4 4 16 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................
B 3/4 4-17 n
l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATOR$..............................................
B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS...............................
B 3/4 5-1
[
3/4.5.4 REFUELING WATER STORAGE TANK......................
B 3/4 5-2 t
3/4.6 CONTAINMENT SYSTEMS l
3/4.6.1 PRIMARY CONTAINMENT.......................................
B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................
B 3/4 6 4 3/4.6.3 CONTAINMENT ISOLATION /ALVES..............................
B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L...................................
B 3/4 6 4 3/4.6.5 ICE CONDENSER.............................................
B 3/4 6 5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.............................................
B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........
B 3/4 7-2 i
3/4.7.3 COMPONENT COOLING WATER SYSTEM............................
B 3/4 7-3 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM..............................
B 3/4 7-3 3/4.7.5 STANDBY NUCLEAR SERVICE WATER P0ND........................
B 3/4 7-3 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM......................
B 3/4 7-3 3/4.7.7 AUXILIARY BUILDING FILTERED EXHAUST SYSTEM................
B 3/4 7-4 3/4.7.8 SNUBBERS..................................................
B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION...............................
B 3/4 7-6 L
3/4.7.10 FIRE SUPPRESSION SYSTEMS..................................
B 3/4 7-6 3/4.7.11 FIRE BARRIER PENETRATIONS.................................
B 3/4 7-7 3/4.7.12 GROUNDWATER LEVEL.........................................
B 3/4 7-7 3/4.7.13 STANDBY SHUTDOWN SYSTEM...................................
B 3/4 7-8 CATAWBA - UNITS 1 & 2 XIV Amendment No. 73 (Unit 1)
Amendment No. 67 (Unit 2)
BASES t
SECTION PAGE 3/4.0 APPLICABILITY,................,,,...,.......................
B 3/4 0 1 i
3/4.1 REACTIVITY CONTROL SYSTEMS I
i 3/4.1.1 BORATION CONTR0L......................,...................
B 3/4 1-1 3/4.1.2 BORATION SYSTEMS..........................................
B 3/4 1 2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................
B 3/4 1 3 t
3/4.2 POWER DISTRIBUTION LIMITS.............................
B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE.....................................
B 3/4 2 1
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3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR CHANNEL FACT 0R......................ENTHALPY RISE HOT l
B 3/4 2 2
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FIGURE B 3/4.2 1 TYPICAL INDICATED AXIAL FLUX THERMAL P0WER.........................OIFFERENCE VERSUS B 3/4 2 3 3/4.2.4 QUADRANT POWER TILT RATI0.................................
B 3/4 2 5 3/4.2.5 DNB PARAMETERS............................................
B 3/4 2-6
^
3/4.3 INSTRUMENTATION r
3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...............
B 3/4 3 1 j
3/4.3,3 MONITORING INSTRUMENTATION................................
B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECT!0N............................
B 3/4 3-7 t
r i
3/4.4 REACTOR COOLANT SYSTEM i
i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............B 3/4 4-1 3/4.4.2 SAFETY VALVES.............................................
B 3/4 4-1 k
3/4.4.3 PRES $URIZER...............................................
B 3/4 4 2 l
3/4.4.4 aEtIEr vAtvES.............................................
B 3/4 4-2 l
3/4.4.5 STEAM GENERATOR $.........................................
B 3/4 4-2 i
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................
B 3/4 4-3 3/4.4.7 CHEMISTRY.................................................
B 3/4 4 5 i
3/4.4.B SPECIFIC ACTIVITY.........................................
B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............................
B 3/4 4 7 I
l CATAWBA - UNITS 1 & 2 XIII l
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q TABLE 4.4 4 (Continued)
TABLE NOTATIONS l
l Until the specific activity of the Reactor Coolant System is restored j
within its limits.
- $ ample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since rcactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
- A gross radioactivity analysis shall consist'of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides l-with half-lives less than 10 minutes and all radioiodines.
The total specific activity shall be the sum of the degassed beta gamma activity and the total-of all identified gaseous activities in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> af ter the sample is taken and extrapolated back to when the sample was taken.
Dete r-mination of the contributors to the gross specific Activity shall be based
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upon those energy peaks identifiable wit 1 a 95% confidence level, The latest available data :tay be used for pure beta-emitting radionuclides.
- A radiochemical analysis for [ shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radiolodines, which is identified in the reactor coolant, The specific activities fjr these individual radio-nuclides shall be used in the determination of :. for the reactor coolant i
sample.
DeterminationofthecontributorstoIshallbebaseduponthose j
energy peaks identifiable with a 95% confidence level.
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CATAWBA - UNITS 1 & 2 3/4 4-31 i
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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSL'.RE/ TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION I
3.4.9.1 The Reactor Coolant System (except the pressuriter) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4 3 during heatup, cooldown, criticality, and inservice leak and i
hydrostatic testing with:
l a.
A maximum heatup of 60'F in any 1-hour period, l
i b.
A maximum cooldown of 100'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, and A maximum temperature change of less than or equal to 10*F in any t
c.
1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY:
At all times, t
ACTION:
[
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation 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 reduce the T,yg and pressure to less than 200'F and 500 psig,
(
respectively, 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
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4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, i
as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4-5.
The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.
l CATAWBA - UNITS 1 & 2 3/4 4-32 Amendment No. 73 (Unit 1)
Amendment No. 67 (Unit 2)
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C 1500 e
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1250 -
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l 1000 =
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750 i
l' CRITICALITY L MIT BASl!D i
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ON INSERVICE HYDROSTATIC l TEST TEMP. (2450F) FORiTHE 250 SERVICE PERIOD UP TO li0 EFPY 3
ACCEPTABLE i
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OPE R ATION i
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0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (CEG. F) l L
CURVI APPLICA8Lt Fom Mt ATUP MAf t RIAL Sallt R Attl UP TO te*P/MM POR THE CONTROLLING MAf TRIAL-UNIT 2 INTimMIDIAf t SHELL limVICE Pt As00 UP TO 10 88PY PLAffD0006-2 l
CONT AIN8 MAmolN of 10*F COPPI A CONTINT-4 Ofwe%
AND to Psic tom PontisLE NICKt L CONTENT. O ti e %
4NETRUMINT t hnoms RTNDT3NII'AL*33
RTwotAP T E R 10 l f PY 1/47, ita't St at, tr'P i
FIGURE 3.4-2 CATAWBA NUCLEAR STATION, REACTOR COOLANT SYSTEM, HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY CATAWBA-UNITS 1 AND 2 3/4 4-33 Amendment No. 73 (Unit 1)
Amendment No. 67 (Unit 2)
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g 1500 i
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1250 l
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I ACCEPTABLE 2
RATES l OPER ATION l
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IN0lCATED TEMPERATURE (DEG, F)
Cymyt APPLiCA4Lt P0m COOLDoveN e4Af f AIAL $ A$18
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R Atts UP 'O its*P/wm 70m THE COWTROLLloso MAf tklAL.VNif 2 INTthestDIATI sHELL i
siavict Pt mooO Up to to t**Y PLAT senos.:
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C04T Aiks edAmeses of Sp*'
COPPt R CONTINT-4 Ofwt%
Awo ao psio eom PoesieLt wickIL CONTE wT-0 91 wis th5T hutat hf I mm0ml.
NTN0TikiflAL-33*F
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hTwotAf f tm 10 t PPY 1/4T,104*p 3/47, tr*P
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FIGURE 3.4-3 CATAWBA NUCLEAR STATION, l
RE ACTOR COOLANT SYSTEM, COOLDOWN LIMITATIONS i
APPLICABLE FOR THE FIRST l
10 E FPY j
CATAWSA-UNITS 1 AND 2 3/4 4 - 34 Amendment No. 73 (Unit 1)
Amendment No. 67 (Unit 2)
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TABLE 4.4-5 C
g REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE c
UNIT 1 UNIT 2 3 CAPSULE VESSEL LEAD LEAD M NUpBER LOCATION FACTOR FACTOR WITHORAWAL TI9E (EFPY) e.
U 58.5' 3.85 4.01 Standby u
V 61*
3.65 3.74 9
W 121.5*
3.85 4.01 Standby X
238.5' 3.85 4.01 Standby Y
241' 3.65 3.74 5
w1 Z
301.5*
3.85 4.05 First Refueling fE
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- 3 en
.O O 22 AA 00
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y REACTOR COOLANT SYST('M p
BAS [$
i 3/4.4.9 FRESSURE/ TEMPERATURE LIMifs The temperature and pressure changes during bestup and cooldown are limited to be consistent with the requirements given in the ASM[ Boiler and Pressure Vessel Code, Section 112, Appendix G:
1.
The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressuriter) shall be limited in accordance with Figures 3.4 2 and 3.4 3 for the service period specified thereon:
I
' Allowable combinations of pressure and temperature for specific a,
temperature change rates are below and to the right of the limit lines shown.
t.imit lines for cooldown rates between those presented may be obtained by interpolation; and b.
Figures 3.4 2 and 3.4 3 define limits to assure prevention of non ductile failure onl For normal operation, other inherent plant characteristics, e.g., y. pump heat addition and press"izer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure temperature ranges.
2.
These limit lines shall be calculated periodically using methods provided
- below, i
3.
The secondary side of the steam generator must not be pressurized above i
200 psig if the temperature of the steam generator is below 70'F, 4.
The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200'F/h, respectively, ard 5.-
System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler 1
i and Pressure Vessel Code,Section XI.
r The fracture toughness properties of the vessel are determined in l
accordance with the 1971 Winter Addenda to Section !!! of the ASME Boiler and Pressure vessel Code and the NRC Branch Technical Position MTEB $*2, and in
)
.accordance with additional reactor vessel requirements.
These properties are then evaluated in accordance with Appendix G of the 1971 Winter Addenda to I
Section !!! of the ASME Boiler and Pressure vessel Code.
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. CATAWBA - UNITS 1 & 2 B 3/4 4-7 t.
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BASES PRESSURE / TEMPERATURE LIMITS (Continued) o Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of the effective full power years (EFPY) of service life as indicated on the applicable heatup or cooldown curves.
The service life period is chosen such I
that the limiting RT at the 1/4T location in the core region is greater NDT than the RT of the limiting unitradiated material.
The selection of such NDT L
a limiting RT assures that all components in the Reactor Coolant System NDT will be operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1.
Reactor opera-t' tion and resultan; fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, based NDT.
t.pon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ARTNDT' For Unit 2 the adjusted reference temperature has been computed using the guidance of Regulatory Guide 1.99, Revision 2.
For Unit 1, the analysis documented in WCAP 11527 and reanalyzed using the guidance of Regulatory Guide 1.99, Revision 2, indicates the heatup and cooldown limit curves in Figures 3.4-2 and 1 4-3 are applicable to both units to predict the shift in RT at the end of the identified service life.
NDT Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available.
Capsules will be removed in accordance with the requirements of l
ASTM E185-82 and 10 CFR Part 50, Appendix H.
The surveillance specimen with-l l
drawal schedule is shown in Table 4.4-5.
The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule l
and the inner wall of the pressure vessel.
Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage l
to the pressure vessel material by using the lead factor and the withdrawal time of the capsule.
The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated NDT l
ART for the equivalent capsule radiation exposure.
NDT Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Cod 9 as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.
1 CATAWBA - UNITS 1 & 2 8 3/4 4-8 Amendment No. 73 (Unit 1)
Amendment No. 67 (Unit 2)
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EFFECTIVE FULL POWER (YEARS) i 1
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FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1Mev) AS A FUNCTION OF FULL POWER SERVICE LIFE CATAWBA - UNITS 1 & 2 B 3/4 4-11 p
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CATAWBA - UNITS 1 & 2 B 3/4 4-12 Amendment No. 73 (Unit 1) l Amendment No. 67 (Unit 2) 1 l
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