ML20085F874
| ML20085F874 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/08/1995 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | Southern Nuclear Operating Co, Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20085F877 | List: |
| References | |
| NPF-68-A-087, NPF-81-A-065 NUDOCS 9506190448 | |
| Download: ML20085F874 (21) | |
Text
-
d# "%cg lt UNITED STATES 3"
f g
NUCLEAR REGULATORY COMMISSION g
WA8HINGTON, D.C. 20666-0001 p
%,...../
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA V0GTLE ELECTRIC GENERATING PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No. NPF-68 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Facility Operating License No. NPF-68 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 3,1994, as supplemented by letter dated March 1, 1995, 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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activitics 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 inimtcal t> 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.
m P
i
. 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-68 is hereby amended to read as follows:
r Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 87
, 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 from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION D
/e?,/qcs.-(~l/l/a..[-
-Werbert 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:
June 8, 1995
pocco Jt UNITED STATES
- g.
- j j
NUCLEAR REGULATORY COMMISSION
^,
WASHINGTON, D.C. 20066-0001 o
GEORGIA POWER COMPANY i
OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA V0GTLE ELECTRIC GENERATING PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No. NPF-81 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Facility Operating License No. NPF-81 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 3, 1994, as supplemented by letter dated March 1, 1995, 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 activi. ties 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.
I
, 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-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.
65, 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 from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION q
Mieblert N. Berkow, Director cf E bAvvl w w
i Project Directorate 11-2 I
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
June 8, 1995
i ATTACHMENT TO LICENSE AMENDMENT NO. 87 FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 A!E TO LICENSE AMENDMENT NO.65 FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 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 Index XVI Index XVI 3/4 4-31 3/4 4-31 3/4 4-31a 3/4 4-31a 3/4 4-32 3/4 4-32 3/4 4-32a 3/4 4-32a
!./4 4-35 3/4 4-35 3/4 4-35a 3/4 4-35a 3 3/4 4-8 8 3/4 4-8 B 3/4 4-9 B 3/4 4-9 B 3/4 4-9a B 3/4 4-9a B 3/4 4-10 B 3/4 4-10 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 8 3/4 4-11 B 3/4 4-14 8 3/4 4-12 B 3/4 4-15 B 3/4 4-13 B 3/4 4-16 B 3/4 4-14 B 2/4 4-17 B 3/4 4-15 i
i i
INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.......... B 3/4 4-1 3/4.4.2 SAFETY VALVES.......................................... B 3/4 4-2 3/4.4.3 PRESSURIZ E R............................................ B 3 /4 4-2 3/4.4.4 RE L I E F VklV E S.......................................... B 3 / 4 4-3 3/4.4.5 STEAM GENERATORS....................................... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE......................... B 3/4 4-4 3/4.4.7 CHEMISTRY.............................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY...................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................ B 3/4 4-7 TABLE B 3/4.4-la UNIT 1 REACTOR VESSEL TOUGHNESS............... B 3/4 4-9 TABLE B 3/4.4-lb UNIT 2 REACTOR VESSEL TOUGHNESS............... B 3/4 4-9a 3/4.4.10 STRUCTURAL INTEGRITY...............................
B 3/4 4-14 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.......................
B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS........................................... 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 1
1 V0GTLE UNITS - 1 & 2 XVI Amendment No. sg (Unit 2)
Unit 1)
Amendment No. o
(
2500 Leak Test Umit <N l,l 2250 I I I 2000 7
1750 f
unacceptasse Operation l
- 1500 1
.I 1250 f
Heatup a e up to N Acceptable Operation g
j 1000 Heatup Rate up tos 1
100'F/hr f
Criticality Umit Based
[
on inservim Hydrostatic TestTemperamre 500 f
/
(246'F) for the Service -
/
Period up to 16 EFPY 250 0
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature ('F)
MATERIAL BASIS Copper Corsent:
Amewned - NAWT%
(Adual 0.083 WT%)
RTNDTW.
Amaned NAT (Actual-207)
RTNDTAt16 EFPY: @ 1/4T = 100.7T
@ 3/4T = 84.11 Figure 3.4-2a Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates up to 100*F/hr). Applicable for the First 16 ETPY (With Margins of 10'T and 60 psig for Instrumentation Errors and Margin of *l4 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region).
Amendment No. 87 (Unit 1) voc:LE UNI:s 1 & 2 3/4 4-31 Amendment No. 65 (Unit 2)
2500 LeakTest Umit Q 2250 2000 i
Unacceptable Operation
)
1750 I J) l 1500 Festu Rate Acceptable up to O'F/hr
}
Operation p
l g
1250 E
o.
1000 Heatup Rate up to 100*F/hr 750 f
/
Criticality Umit Based
/
on inservios Hydrostatic Test Temperature 500 f256'F) for the Service -
b I eriod up to 16 EFPY a
250 0
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (*F)
MATERIAL BASIS Copper Content Assumed NAWT%
(Actual 0.05 WT%)
NDT nM
- h. NA T RT i
(Actual 50T)
At 16 EFPY: @ 1/4T 112T RTNDT
@ 3/4T 94T Figure 3.4-2b Unit 2 Reactor Coolant System Heatup Limitations (Heatup rater, up to 100*F/hr) Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Wessel Beltline Region).
l VOGTLE UNITS 1 & 2 3/4 4-31a Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
2500 l
2250 f
Unacceptable Operation l
1750
- 1500 e
g 1250 E
Acceptable Operation E
1000 E
l 750 Cooldown Rates 'F/hr 500 3
250 1Luv 0
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature ('F)
MATERW. BASIS Copper Content Assuned NAWT%
(Adual 0.083 WT%)
RTNDT nk Assumed-NA Y I
(Actual-20T)
RTNDTAt 16 EFPY: @ 1/4T = 100.7T
@ 3/4T = 44.1T Figure 3.4-3a Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100*r/hr)
Applicable for the First 16 EFPY (With Margins of 10*r and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor vessel Beltline Region).
voGTLE tmITs 1 & 2 3/4 4-32 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
1 l
2500 l
2250 2M0 1750 l
Unacceptable Operation 1500 1250 dE Acceptable Operation
]
1000 750 Cooldown F
Rates 'F/hr 500 a
250 gg /
0 0
50-100 150 200 250 300 350 400 450 500 Indicated Temperature ('F) 1 MATERIAL BASIS Copper Content Assumed NAWT%
(Adual 0.06 WT%)
RTNOTN Assumed NAT (Amual sot)
RTNDTAt 16 EFPt 9 1/4T = 112T
@ 3/4T 941 Figure 3.4-3b Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100'F/hr)
Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region).
VOGTLE UNITS 1 & 2 3/4 4-32a knendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
800 780 760 740 (212,726)
(350,726) 720 700 680 3f 660 f
640 620 600
{
580 J
560 I 0'"I 540 520 500 50 100 150 200 250 300 350 TRTD - Auctioneered Low Measured RCS Temperature ('F)
Figure 3.4-4a Unit 1 Maximum Allowable Nominal PORV Setpoint for the Cold overpressure Protection System voG;:.E UNITS 1 & 2 3/4 4-35 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
800 780 760 y 740 (222,726)
(350,726)
S 720 l80
/
6 660 f
640 I
620 600 j
580 1560 g
540
/
I 520 (70,516) 500 50 100 150 200 250 300 350 TRTD - Auctioneered Low Measured RCS Temperature (*F)
Figure 3.4-4b Unit 2 Maximum Allowable Nominal PORV Setpoint for the Cold Overpressure Protection System vos:LE UNITS 1 & 2 3/4 4-35a knendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
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.
-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 Pressure Vessel Code,Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-82, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendices, " Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," 1986 Edition 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 for Unit I and Figures 3.4-2b and 3.4-3b for Unit 2 are applicable for up to 16 EFPY and were developed based on the actual material properties of the most limiting material. The most limiting material are shown in Table B 3/4.4-la for Unit 1 and Table B 3/4.4-lb for Unit 2.
Heatup and cooldown limit curves are calculated using the most limiting valueofthenil-ductilityreferencetemperature,RT[e,EFPYservicelife at the end of the Effective Full Power Years (EFPY) of service life. D period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT ofthelimNingunirradiatedmaterial. The selection of such a limiting k assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.
V0GTLE UNITS - 1 & 2 8 3/4 4-8 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
TABLE B 3/4.4-la UNIT 1 REACTOR VESSEL TOUGHNESS A,
r E
C COMP CU NI INITIAL 16 EFPY RTNOT f
COMPONENT faDI 111 ill L i'El 1/4-tf*F)
'3/4-tf*F)
Closure Head Flange 0.70 20
[
Vessel Flange 0.71 0
Intermediate Shell B8805-1
.0.083 0.597 0
80.7 64.1 Intermediate Shell*
B8805-2 0.083 0.610 20 100.7
~
'84.1 Intermediate Shell B8805-3 0.062 0.598 30 97.5 76.4 Lower Shell B8606-1 0.053 0.593 20 77.6 57.6 Lower Shell B8606-2 0.057 0.600 20 81.9 62.5 Lower Shell B8606-3 0.067 0.623 10 80.8 60.6 R
Circ. Weld 101-171 0.039 0.102
-80
-21.7
-39.9 da Long. Weld 101-124A 0.039 0.1'02
-80
-31.8
-48.5 Long. Weld 101-124B 0.039 0.102
-80
-30.0
-47.0 g{3 Long. Weld 101-124C 0.039 0.102
-80
-30.0
-47.0 Long. Weld 101-142A 0.039 0.102
-80
-30.0
-47.0 mgg Long. Weld 101-142B 0.039 0.102
-80
-31.8
-48.5 gg Long. Weld 101-142C 0.039 0.102
-80
-30.0
-47.0 8:ss EE OX S ".
- Limiting material i
. m
,,., ~.....,,... -...,,,,,,.. -
TABLE B 3/4.4-lb UNIT 2 REACTOR VESSEL TOUGHNESS 8
r I
- 3 COMP CU NI INITIAL 16 EFPY RTNOT COMPONENT EDIlE L%1 L%1 g
1/4-t(*F) 3/4-tf*F) l" Closure Head Flange 0.72 10 Vessel Flange 0.87
-60 Intermediate Shell R4-1 0.06 0.64 10 81 62 Intermediate Shell R4-2 0.05 0.62 10 72 54 Intermediate Shell R4-3 0.05 0.59 30 92 74 Lower Shell B8825-1 0.05 0.59 40 102 84 j$
Lower Shell R8-1 0.06 0.62 40 111 92 Lower Shell*
B8628-1 0.05 0.59 50 112 94 Circ. Weld 0.06 0.12
-30 55 31 l[j[
Long. Weld 0.07 0.13
-10 83 56 oo kk EE
.F F 2
E?i?
- 3. 3.
se ga -
- Limiting material
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
The reactor vessel materials have been tested to determine their initial RT
- the results of these tests are shown for Units 1 and 2 in Tables Bk4.4-laandb,respectively.
Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, based upon the fluen r.ce, copper content, and nickel content of the material in question, can be predicted using Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." 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 16 EFPY as well as adjustments for possible errors in the p., at the end of (Unit 2) include predicted adjustments for this shift in RT ressure and temperature sensing instruments of 60 psig and 10'F, respectively.
In addition, these curves include a pressure adjustment of 74 psig to account for the pressure differential between the wide range pressure transmitter and the belt line region.
from the material., determined in this manner may be used until the results Values of ART 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 withdrawal 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 location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be uscd 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 y
the calculated ART., for The equivalent capsule radiation exposure.
Allowable pressu:m-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section 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) technology.
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 curves 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 i
1 V0GTLE UNITS - 1 & 2 B 3/4 4-10 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2) i
REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY (Continued)
Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through the 1975 Winter Addenda.
3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head, ensures that the capability exists to perform this function.
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737,
" Clarification of TMI Action Plan Requirements," November 1980.
~
1 V0GTLE UNITS - 1 & 2 B 3/4 4-15 Amendment No. 82 (Unit 1)
Amendment No. 65 (Unit 2)
L
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RT is used and this includes the radiation-induced cooldown c.,, correspondi 7,
shift, ART ng to the end of the period for which heatup and urves are generated.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, i for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K,, for i
the metal temperature at that time. K, is obtained from the reference i
fracture toughness curve, defined in Appendix G to the ASME Code. The K,i curve is given by the equation:
K, - 26.78 + 1.223 exp [0.0145(T-RT,7 + 160))
(1) i Where: K, is the reference stress intensity factor as a function of the metal i
temperatureTandthemetalnil-ductilityreferencetemperatureRT*In.
- Thus, the governing equation for the heatup-cooldown analysis is defined Appendix G of the ASME Code as follows:
CK,+K sK, (2) i ir 3
1 Where: K, - the stress intensity factor caused by membrane (pressure) i
- stress, K, - the stress intensity factor caused by the thermal gradients, 3
K, - constant provided by the Code as a function of temperature 3
relative to the RT,7 of the material, C - 2.0 for level A and B service limits, and C - 1.5 for inservice hydrostatic and leak test operations.
At any time during the heatup or cooldown transient, K is determined by the metal temperature at the tip of the postulated flaw, tb,e appropriate value for RT and the reference fracture toughness curve. The thermal stresses y
resultIng, from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K, for the reference flaw is computed.
FromEquation(2)thepressurestressEntensityfactorsare obtained and, from these, the allowable pressures are calculated.
COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of V0GTLE UNITS - 1 & 2 B 3/4 4-11 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2) t
REA0TORCOOLANTSYSTEM{
BASES PRESSURE / TEMPERATURE LIMITS (Continued) the vessel wall. During cooldown, the controlling location of the flaw is always at the inside o1l the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.
From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situa-tion.
It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K, at the 1/4T location i
for finite cooldown rates than for steady-state operation.
Furthermore, if
, the calculated conditions exist such that the increase in K,, exceeds K,lhe steady-state value.
allowable pressure during cooldown will be greater than The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.
HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4T crack during heatup is lower than the K,kure. the 1/4T crack,,during steady-state conditio for at the same coolant tempera During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different;K 's for steady-state and finite heatup rates do not.
offset each other and the, pressure-temperature curve based on steady-state i
conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value j
of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
i V0GTLE UNITS - 1 & 2 B 3/4 4-12 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each heatup rate of interest must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Next, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
Finally, the new 10 CFR 50 Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange regions is considered.
This rule states that the minimum metal temperature of the closure flange regions should be at least 120*F higher than the limiting RT for these er regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Unit 1).
For Unit I the minimum temperature of the closure flange and vessel flange regions is 140'F, since the limiting RT is er 20*F (see Table B 3/4-4.la).
For Unit 2, the minimum temperature of the closure flange and vessel flange regions is 130*F, since the limiting RT, is 10*F (Table B 3/4-lb). Thesevaluesincludemarginof10*Fand60psigIor instrumentation errors. The heatup and cooldown curves as shown in Figures 3-4.2a and 3-4.3a for Unit I and the heatup and cooldown curves as shown in Figures 3-4.2b and 3-4.3b for Unit 2 are impacted by the new 10 CFR 50 rule.
Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limite are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
V0GTLE UNITS - 1 & 2 8 3/4 4-13 Amendment No. 87 (Unit 1)
Amendment No. 65 (Unit 2)
l REACTOR COOLANT SYSTEM BASES COLD OVERPRESSURE PROTECTION SYSTEMS The OPERABILITY of two PORVs, two RHR suction relief valves, a PORV and RHR SRV, or an RCS vent capable of relieving at least 670 gpm water flow at 470 psig ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350*F. The PORVs have adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures, or (2) the start of all three charging L
pumps and subsequent injection into a water-solid RCS. The RHR SRVs have adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle RCP with the secondary to primary eter temperature difference of the steam generator less than or equal to 25'F v an RCS temperature of 350'F and varies linearly to l_
50*F at an RCS temperature of 200*F or less, or (2) the start of all three charging pumps and subsequent injection into a water-solid RCS. A combination j
of a PORV and a RHR SRV also provides overpressure protection for the RCS.
The Maximum Allowed PORV Setpoint for the Cold Overpressure Protection System (COPS) is derived by analysis which models the performance of the COPS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that the nominal 16 EFPY Appendix G reactor vessel NDT limits criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur Technical Spec-ifications require lockout of all safety injection pumps while in MODES 4, 5, and 6 with the reactor! vessel head insiclied and disallow start of an RCP if secondary temperature is more than 50*F above primary temperature. Additional temperature limitations are placed on the starting of a Reactor Coolant Pump in Specification 3.4.1.3.1 These limitations assure that the RHR system remains i
within its ASME design limits when the RHR relief valves are used to prevent RCS overpressurization, The Maximum Allowed PORV Setpoint for the COPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accor-dance with the schedule in Table 16.3-3 of the VEGP FSAR.
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3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, I
and 3 components ensure that the structural integrity and operational readiness l
of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the i
ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
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V0GTLE UNITS - 1 & E B 3/4 4-14 Amendment No. 87 (Unit 1) i Amendment No. 65 (Unit 2) g