ML20058C413
| ML20058C413 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/19/1990 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20058C417 | List: |
| References | |
| NPF-68-A-036, NPF-81-A-016 NUDOCS 9011010219 | |
| Download: ML20058C413 (12) | |
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UNITED STATES
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e GEORGIA POWER-COMPANY OGLETHORPE POWER CORPORATION HUNICIPAL ILECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA V0GTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 i
License No. NPF-68 1.
The Nuclear Regulatory Comission (the Commission) has found that:
(
A.
The app (lication for amendment to the Vogtle Electric Generating Plant, i
Unit 1 the facility), facility Operating License No. NPF-68 filed by.
the Georgia Power Company,. acting for itself, Oglethorpe Power Corpo-j ration, Municipal Electric Authority of Georgia, and City of' Dalton, Georgia (the licensees), dated July 28, 1989, complies with the i
standards and requirements of the Atomic Energy Act of-1954, as amended '(the Act), and the Comission's rules and regulations set -
forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the.
j provisions of the Act, and the rules and regulations of the 1
Comission;-
i 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 Comission's regulations set forth in 10 CTR Chapter I; i
D.
The issuance of this license amendment will not'be inimical to the,
'I comon 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 Comission's regulations and all applicable requirements have-been satisfied.
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2 2.
Accordingly, the-license is hereby amended by page changes to.the Technical i
Specifications as indicated in the attachment to this. license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby amended to read as follows:
Technical Specifications and Environmentel Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 36, and the Environmental Protection Plan.
contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATOR \\ COMMISSION
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David B.'Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation-
Attachment:
Technical Specification Changes Date of Issuance:
October 19, 1990 1
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UNITED $TATES
'o NUCLEAR REGULATORY COMMISSION j
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GEORGIA POWER COMPANY.
P OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA
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CITY OF DALTON, GEORGIA V0GTLE ELECTRIC GENERATING PLANT, UNIT 2 3
AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.16 License No. NPF-81 1.
The Nuclear Regulatory Comission-(the Comission) has found that:
The app (lication for amendment to the Vogtle Electric Generating Pla A.
i Unit 2 the facility), Facility Operating License No.. NPF-81~ filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of. Dalton, Georgia (thelicensees),datedJuly 28, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the.
provisions of the Act, and the rules and regulations of the Comission; C.
Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and-(ii) that such. activities will be conducted in compliance with the Comission's; regulations set forth in 10 CFR Chapter I; D.
The issuance of this license amendment 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 amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amenoment t
and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby
~
amended to read as follows:
Technical Specifications and Environmental-Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 16, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
/
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a'.cc'U David B. Matthews, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
l Technical Specification Changes Date of Issuance:
October 19, 1990 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 36-FACILITY OPERATING LICENSE NO. NPF-68 AND LICENSE AMENDMENT NO.16 FACILITY OPERATING LICENSE NO. NPF-81 DOCKETS N05.'50-424 AND 50-425 keplace 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. The corresponding overleaf pages are also provided to saintain doct! ment completeness.
Amended Page Overleaf Page 3/4 4-31 3/4 4-32 3/4 4-31a B 3/4 4-8 8 3/4.4-7
(
B 3/4 4 B 3/4 4-9a i
L 1
4 l
UNIT 1 3000 l
CURVE APPLICABLE FOR THE SERVICE PERIOD UP TO 13 EFPY l
~f, a.1/4 T < 110*F l
- b. 3/4 T< 87'F LEAK TEST l
LIMIT l
CRITICALITY i
LIMIT FOR 60'F/hr -
2000 UNACCEPTABLE
~
HEATUP OPERATION e*
CRITICALITY j
LIMIT f
FOR 100'F/hr ~
HEATUP.-
60'F/hr HEATUP 0 1000 CURVE /
/
. BASED ON INSERVICE '
[
HYDROSTATIC TEST E
r TEMPERATURE (255'F) l p,.
t F
FOR THE SERVICE PERIOD UP TO 13 EFPY 100'F/hr kHEATUP ACCEPTABLE CURVE OPERATION I
0.0 0.0 100 200 300 400 500 LOWEST INDICATED RCS Toold TEMPERATURE ('F)
MATERIAL BASIS Coppet Content. Assumed. 0.10 Wt %
(Astuel. 0.00 Wt %)
E T nittel:
Assumed. 40'F -
RT I
(Actuoi. 30'F)
RT8 TDT After 13 EFPY e 1/4 T *110'F e S/4 T et?'F.
FIGURE 3.4-2a UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 13 EFPY V0GTLE UNITS 1 & 2 3/4 4-31 Amendment No. 36 (Unit 1)
Amendment No. 16 (Unit 2)
\\
UNIT 2 CURVE APPLICABLE FOR THE SERVICE PERIOD UP TO 16 EFPY f
LEAK TEST
{2000 LIMIT f, f CRITICALITY LIMIT FOR 1
800F/hr HEATUP l
UNACCEPTABLE E
OPERATION g 1500 0
80 F/hr HEATUP CURVE CRITICALITY LIMIT FOR 0
100 F/hr HEATUP Z
5 a
)
ACCEPTABLE OPERATION E 1000 100*F/hr HEATUP CURVE %
"a BASED ON I SERVICE g
HYDROSTATIC TEST M
TEMPERATUR E (268'F)
FOR THE SERVICE PERIOD UP TO 18 EFPY
'O.0 100-200 300 400 500 LOWEST INDICATED RCS Tg TEMPERATURE ('F) l MATERI AL BASIS Cooper Cententi Asemmes. 0.10 Wt %
( Asteel. 0.06 wt %)
ATNOT
- Thaf g
p RT Ahor it EFPY e 1/4 T = 123*F NOT e 3/4 7 = 97 F FIGURE 3.4-2b UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO~16 EFPY V0GTLE UNITS - 1 & 2 3/4 4-31a y
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UNIT 1 3000 l -
CURVE APPLICABLE FOR THE SERVICE PERIOD UP TO 13 EFPY '
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2000 m
UNACCEPTABLE OPERATION ACCEPTABLE OPERATION iu A
$ 1000 COOLDOWN p
RATE 5
('F/hr) g 0sV/
E 60 /
100 0.0 0.0 -
100 200 300 400 500 LOWEST INDICATED RCS Toold TEMPERATURE (*F)
PATTRIAL SAll$
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%)
RTMOT W:
M.40*F (Astuel. 30*P)
E Ab U N 1 "INDT l
FIGURE 3.4-3a UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 13 EFPY V0GTLE-UNITS - 1 & 2 3/4 4-32 Amendment No. 36 (Unit 1)
Amendment No.16 (Unit 2)
l.
-1 REACTOR COOLANT SYSTEM BASES l
l SPECIFIC ACTIVITY (Continued)
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to per-form the sampling, transport the sample, and perform the analysis of about 90 minutes.
After 90 minutes, the specific count should be made for gases 1
(i.e., xenons and kryptons) and particulates (i.e., cobalt and cesiums) in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties.
The counter should be reset to a reproducible efficiency versus energy.
It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple count-ing of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about I day, about I weak, and about 1 month.
The identification of 95% of the grgss specific activity by definition does not obligate VEGP into calculating E every time gross activity is determined.
Reducing T to less than 500 F prevents the release of activity should asteamgenerat@9 tube rupture since the saturation pressure of the reactor coolant is below the lif t pressure of the atmospheric steam relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are.
limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:
1.
The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall-be limited in accordance-l with Figures 3.4-2a and 3.4-3a (Unit 1), Figures 3.4-2b and 3.4-3b (Unit 2) l for the service period specified thereon:
a.
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit-lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation; and b.
Figures 3.4-2a and 3.4-3a (Unit 1), Figures 3.4-2b and 3.4-3b (Unit 2) define limits to assure' prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
V0GTLE UNITS - 1 & 2 B 3/4 4-7
' REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) 2.
These limit lines shall-be calculated periodically using methods provided
- below, 3.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F, 4
4.
The pressurizer heatup and cooldown rates shall not exceed 100*F/h and 200'F/h, respectively.
The auxiliary spray shall not be used if the temperature difference between the pressurizer and the auxiliary spray
' fluid is greater than 625'F, and 5.
System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pra sure Vessel Code,Section XI.
The 9acture toughness properties of the ferritic materials in the reactor vessel rce determined in accordance with the NRC Standard Review Plan, ASTM E185-87, and in accordance with additional reactor vessel requirements.
These properties are then evaluated in accordance with Appendix G of the 1972 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.
The heatup and cooldown limit curves shown in Figures 3.4-2a and 3.4-3a are applicable to Unit 1 for up to 13 EFPY and are based on Westinghouse-l developed generic curves which were developed assuming a 40'F initial RT and a copper content of 0.10 WT% for the most limiting material.
These b ves are applicable to Unit I since its most limiting material (Table B 3/4.4-la) has both a lower initial RTNDT (30*F) and a lower copper content (0.06 WT%).
These curves, however, are not applicable to Unit 2, since its most limiting material (Table B 3/4.4-lb) has a higher initial RT (50 compared to 40*F).
Separate heotup and cooldown limit curves w b developed based on the actual material properties of the most limiting material for Unit 2 up to 16 EFPY.
The Unit 2 curves are shown in Figures 3.4-2b and 3.4-3b.
Heatup and cooldown limit curves are calculated using the most limiting vclua of the nil-ductility reference temperature, RTNDT, at the end of the Effective Full Power Years (EFPY) of service life.
The EFPY service life period is chosen such that the limiting RT at the 1/4T location in the core region-NDT l
is greater than the RT f the limiting unirradiated material.
The selection NDT l
of such a limiting RT assures that all components in the Reactor Coolant NDT System will be operated conservatively in_accordance with applicable Code requirements.
l V0GTLE UNITS - 1 & 2 8 3/4 4-8 Amendment No. 36 (Unit 1)
Amendment No.16 (Unit 2) l
f TABLE B 3/4.4-1b Es y}
. UNIT 2 REACTOR VESSEL TOUGHNESS c.
ASME AVERAGE T
RT 5
COMP MATERIAL CU Ni P
NOT NOT NMWO*
j 51
. COMPONENT-
-C00E TYPE
(%)
1%1 1%}
(*F)
('F)
(FT-LB) j Closure Head Dome R9-1 A5338 CL. 1 0.07 0.61 0.008
-40
-30 123 Closure Head Torus R10-1 A5338 CL. 1 0.07 0.64 0.010
-30 0
84 i
Closure Head Flange R7-1 A508 CL. 2 0.72 0.011 10 10 130 n,
Vessel Flange R1-1 A508 CL. 2 0.87 0.011
-60
-60 115
^
Inlet Nozzle B9806 A508 CL. 2 0.07 0.84 0.010
-50
-50
- 119 Inlet Nozzle B9806-2 A508 CL. 2 0.06 0.83 0.009
-40
-40 128 Inlet Nozzle RS A508 CL. 2 0.09 0.87 0.008
-20
-20 147 Inlet Nozzle-RS-2 A508 CL. 2 0.08 0.85 0.009 -20 134 l
Outlet Nozzle R6-3 A508 CL. 2
. 0.69 0.011
-10
-10 122 Outlet Nozzle R6-4
'A508 CL. 2 0.66 0.010
-10
-10 140 l
Outlet Nozzle B9807-3 A508 CL. 2 0.66 0.005
-30
-30 116 R3 Outlet' Nozzle.
B9807-4:
A508 CL. 2 0.64:
0.010 10~
10
. 132
-i d*
Nozzle Shell
-R3-1 A5338 CL. 1.
0.20 0.67 0.015 0
20 79
' t*
Nozzle Shell R3-2 A5338 CL. 1 0.20 0.67 0.015 0
40 79
. g?
Nozzle Shell
-R3-3:
A5338 CL. 1 0.15 0.62 0.010
~-10 60 84 Intermediate Shell R4-1 A5338 CL. 1 0.06 0.64 0.009
-20 10 95 Intermediate Shell R4 A5338 CL. 1.
0.05 0.62 0.009 10 104 Intermediate-Shell R4-3 A5338 CL. 1 0.05 0.59.
0.009 0
30 84 Lower Shell 88825-1 A5338 CL. 1 0.05 0.59 0.006
-20 40 83 Lower Shell R8-1 A5338 CL. 1 0.06 0.62 0.007.
-20 40 87 Lower Shell**
R8628-1 A533B CL. l' O.05 0.59 0.007
-20 50 85 Botton Head Torus R12 A5338 CL. 1 0.17
' O.64 0.012. -20
-20 89 Botton Head Dome R11-1 A5338 CL. 1 0.10 0.62 0.008
-30
-30 115 Intermediate & Lower G1.60 SAW 0.07 0.13 0.007
-10
-10 147 Shell Vertical W ld Seams e
Intermediate to Lower E3.23 SAW 0.06 0.12 0.007'
-50
-30 90 Shell Girth Weld Seam i
- Upper Shelf energy, normal to major working direction.
l
- Limiting material.
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REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown for Units 1 and 2 in Table B 3/4.4-la and b, respectively.
Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT There-fore,anadjustedreferencetemperature,baseduponthefluence,che.r content, and phosphorus content of the material-in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by either Regulatory NDT Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation l
Damage to Reactor Vessel Materials," or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2.
The heatup and cooldown limit curves of Figures 3.4-2a and 3.4-3a (Unit 1), Figures 3.4-2b and 3.4-3b (Unit 2) include predicted adjustments for this shift in RT at the end of 13 (Unit 1) and 16 (Unit 2) l NDT EFPY as well as adjustments for possible errors in the pressure and temperature sensing instrument.s.
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 ASTM E185-82 and 10 CFR Part 50, Appendix H.
The surveillance specimen with-drawal schedule is shown in Table 16.3-3 of the VEGP FSAR.
The lead factor represents the relationship between the fast neutron flux density at the loca-tion of the capsule and the inner wall of the reactor vessel.
Therefore, the results obtained from the surveillance specimens can ce used to predict future radiation damage to the reactor 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 NDT the calculated ART f r 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 Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) tech-nology.
In the calculation procedures a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.
The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.
Therefore, the reactor operation limit cur-ves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.
To assure that the radiation embrittlement effects are accounted for in the calculation of the l
V0GTLE UNITS - 1 & 2 B 3/4 4-10 Amendment No. 36 (Unit 1)
Amendment No.16 (Unit 2)
'