ML20099M005
| ML20099M005 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/28/1995 |
| From: | Matthews D NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20099M006 | List: |
| References | |
| NUDOCS 9601030130 | |
| Download: ML20099M005 (20) | |
Text
-
p Klogk UNITED STATES g
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20666-00N
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]
VIRGINIA ELECTRIC AND POWER COMPAl((
DOCKET NO. 50-280 SURRY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE 1
1 Amendment No. 207 License No. DPR-32 j
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated June 8, 1995, 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; 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.
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 activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part i
51 of the Comission's regulations and all applicable requirements have been satisfied.
1 9601030130 951228 PDR ADOCK 05000280 P
i
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-32 is hereby amended to read as follows:
(B)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 207, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
.,,l
/.
i
/
bh i
David B. Matthews, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 28, 1995
f
~%
UNITED STATES y
g g
NUCLEAR REGULATORY COMMISSION t
WASHINGTON. D.C. 20e864001
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i t
VIRGINIA ELECTRIC AND POWER COMPANY I
DOCKET NO. 50-281 i
SURRY POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 207 License No. OPR-37 4
i j
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power i
Company (the licensee) dated June 8, 1995, 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 i
i 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 2
4 and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; The issuance of this amendment will not be inim'ical to the common u.
defense and security or to the health and safety of the public; and 1
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.
J 4
4
.- m
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-37 is hereby amended to read as follows:
(B)
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 207, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION I
ut David B. Matthews, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 28, 1995
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
207 TO FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO.
207 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Paaes Insert Paaes TS 3.1-3 TS 3.1-3 TS 3.1-6 TS 3.1-6 TS 3.1-9 TS 3.1-9 TS 3.1-10 TS 3.1-10 TS 3.1-11 TS 3.1-11 TS 3.1-12 TS 3.1-12 TS 3.1-18 TS 3.1-18 i
TS 3.1-19 TS 3.1-19 TS 3.1-23a TS 3.1.23a TS 3.1-26 TS 3.1-26 TS 3.1-27 TS 3.1-27
)
TS 3.1-28 TS 3.1-28 TS 3.1-29 TS 3.1-29 Figure 3.1-1 Figure 3.1-1 Figure 3.1-2 Figure 3.1-2 Figures 3.1-3 and 3.1-4
)
TS 3.1-3 e.
When all three pumps have been idle for > 15 minutes, the first pump shall not be started unless: (1) a bubble exists in the pressurizer or (2) the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.
2.
Steam Generator A minimum of two steam generators in non-isolated loops shall be l OPERABLE when the average Reactor Coolant System temperature is greater than 350*F.
3.
Pressurizer Safety Valves a.
Three valves shall be OPERABLE when the head is on the reactor vessel and the Reactor Coolant System average temperature is greater than 350 F, the reactor is critical, or the Reactor Coolant System is not connected to the Residual Heat Removal System.
b.
Valve lift settings shall be maintained at 2485 psig i 1 percent *
- The as-found tolerance shall be i3% and the as-left tolerance shall be i1%.
~
Amendment Nos. 207 and 207
TS 3.1-6 B.
HEATUP AND COOLDOWN Specification 1.
Unit 1 and Unit 2 reactor coolant temperature and pressure and the system heatup r,nd cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figures 3.1-1 and 3.1-2.
Heatup:
Figure 3.1-1 may be used for heatup rates of up to 60'F/hr.
Cooldown:
Allowable combinations of pressure and temperature for specifc' cooldown rates are below and to the right of the limit lines as shown in TS Figure 3.1-2. This rate shall not exceed 100*F/hr.
Cooldown rates between those shown can be obtained by interpolation between the curves on Figure 3.1-2.
Core Operation:
During operation where the reactor core is in a critical condition (except for low level physics tests), vessel metal and fluid temperature shall be meintained above the reactor core criticality limits specified in 10 CFR 50 Appendix G. The reactor shall not be made critical when the reactor coolant temperature is below 522'F as specified in T.S. 3.1.E.
2.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is l
below 70*F.
Amendment Nos. 207 and 207 l
TS 3.1o9 Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, RTNDT, at the end of 28.8 Effective Full Power Years (EFPY) and 29.4 EFPY for Units 1 and 2, respectively. The most limiting value of RTNDT (228.4*F) occurs at the 1/4-T, 0* azimuthal location in the Unit 1 intermediate-to-lower sheli circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RTNDT of the limiting unitradiated material.
This ensures that all components in the Reactor Coolant System will bd l
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 presented in report BAW-2222,
" Response to Closure Letters to NRC Generic Letter 92-01, Revision 1,"
dated June,1994 and are reproduced in Tables 3.1-1 and 3.1-2. Reactor operation and tesultant fast neutron (E greater than 1 MEV) Irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2
- Effects of Residual Elements on Predicted Radiation Damcge to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively (as well as adjustments for location of the pressure sensing instrument).
Surveillance capsules will be removed in accordance with the l
requirements of ASTM E185 82 and 10 CFR 50, Appendix H.
The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated NDT or the equivalent capsule radiation exposure, or when the ART f
service period exceeds 28.8 EFPY or 29.4 EFPY for Units 1 and 2, l
respectively, prior to a scheduled refueling outage.
Amendment Nos. 207 and 207
TS 3.1-10 Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section ill of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.
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 semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of one and one half T 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 lll 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 non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNDT, is used and this includes the radiation induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves 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, for the combined thermal and pressure stresses at any time during i
heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. KIR s obtained from i
the reference fracture toughness curve, defined in Appendix G to the ASME Code. The KIR curve is given by the equation:
4 KIR = 26.78 + 1.223 exp (0.0145(T-RTNDT + 160))
(1) i where KIR s the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RTNDT. Thus, the goveming equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C K[y + K s KIR (2) it where, KIM s the stress intensity factor caused by membrance (pressure) i stress.
Amendment Nos. 207 and 207
i i
l, TS 3.1-11 K is the stress intensity factor cau' sed by the thermal gradients it 1
I l
KIR s provided by the code as a function of temperature relative to i
j the RTNDT of the material.
j C - 2.0 for level A and B service limits, and i
1 C - 1.5 for inservice hydrostatic and leak test operations.
At any time during the heatup or cooldown transient, KIR s determined i
by the metal temperature at the tip of the postulated flaw, the appropriate l
value for RTNDT. and the reference fracture toughness curve. The l
thermal stresses resulting from temperature gradients through the vessel l
wall are calculated and then the corresponding thermal stress intensity l
factor, K, for the reference flaw is computed. From Equation (2) the l
it
)
pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
The heatup limit curve, Figure 3.1-1, is a composite curve which was i
prepared by determining the most conservative case, with either the j
inside or outside wall controlling, for any heatup rate up to 80*F por hour.
l l
The cooldown limit curves of Figure 3.12 are composite curves which l
were prepared based upon the same type analysis with the exception j
that the controlling location is always the inside wall where the cooldown therma! gradients tend to produce tensile stresses while producing i
compressive stresses at the outside wall. The cooldown limit curves are l
valid for cooldown rates up to 100*F/hr. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2,.respectively.. The adjusted reference
)
temperature was calculated using materials properties data from the l
B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) documented in the most recent revision to BAW-1543
)
and reactor vessel neutron fluence data obtained from plant specific j
- analyses, i
Amendment Nos. 207 and 207
TS 3.1-12 P
THIS PAGE HAS BEEN INTENTIONALLY DELETED i
l 1
Amendment Nos. 207 and 207
1 1
i j
TS 3.1-18 i
l E. Minimum Temnerature for Criticality l
Soecifications 4
j 1.
Except during LOW POWER PHYSICS TESTS, the reactor shall not be j
made critical at any Reactor Coolant System temperature above which j
the moderator temperatur's coefficient is more positive than the limit j
specified... the CORE OPERATING LIMITS REPORT. The maximum j
upper limit for the moderator temperature coefficient shall be:
4 l
a.
+ 6 pcm/*F at less than 50% of RATED POWER, or i
b.
+ 6 pcm/*F at 50% of RATED POWER and linearly decreasing to 0 l
pcm/ F at RATED POWER.
l 2.
In no case shall the reactor be made critical with the Reactor Coolant
]
System temperature below the limiting value of RTNDT + 10*F, where the limiting value of RT i
NDT s as determined in Part B of this specification.
i j
3.
When the Reactor Coolant System temperature is below the minimum
)
temperature as specified in E-2 above, the reactor shall be subcritical by
{
an amount equal to or greater than tne potential reactivity insertion due to j
primary coolant depressurization.
l l
4.
The reactor shall not be made critical when the Reactor Coolant System l
temperature is below 522*F.
1 l
Basis 1
l During the early part of a fuel cycle, the moderator temperature coefficient may j
be calculated to be slightly positive at coolant temperatures in the power operating range.
The moderator coefficient will be most positive at the l
)
beginning of cycle life, when the boron concentration in the coolant is the l
greatest. Later in the cycle, the boron concentration in the coolant will be lower j
and the moderator coefficient will be less positive or will be negative in the l
power operating range. At the beginning of cycle life, during pre-operational physics tests, measurements are made to determine that the moderator l
coefficient is less than the limit specified in the CORE OPERATING LIMITS REPORT.
Amendment Nos. 207 and 207
TS 3.1-19 The requirement that the reactor is not to be made critical when the moderator coefficient is greater than the low power limit specified in the CORE OPERATING LIMITS REPORT has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requi,oment is waived during LOW POWER PHYSICS TESTS to permit measurement of reactor moderator coefficient and other physics design parameters of interest.
During physics tests, special operation precautions will be taken. In additio$,
the strong negative Doppler coefficient (2)(3) and the small integrated Delta luk would limit the magnitude of a power excursion resulting from a reduction of moderator density.
The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below the limiting value of RTNDT + 10*F provides l
increased assurance that the proper relationship between Reactor Coolant System pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps.
The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below 522*F provides added assurance that the assumptions made in the safety analyses remain bounding by maintaining the moderator temperature within the range of those analyses.
If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.
(1)
UFSAR Figure 3.3-8 (2)
UFSAR Table 3.3-1 (3)
UFSAR Figure 3.3-9 Amendment Nos. 207 and 207
j i
TS 3.1-23a (3)
During the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, maintain a bubble in the pressurizer with a maximum narrow range level of 33%,
j or (4)
Maintain two Power Operated Relief Valves (PORV)
OPERABLE with a lift setting:of s 390 psig and verify each l j
PORV block valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, j
or j
(5)
The RCS shall be vented through one open PORV or an equivalent size opening as specified below:
(a) with the RCS vented through an unlocked open vent
{
path, verify the path is open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or (b) with the RCS vented through a locked open vent path verify the path is open at ivast once per 31 days.
2.
The requirements of Specification 3.1.G.1.c.(4) may be modified as follows:
a One PORV may be inoperable in INTERMEDIATE SHUTDOWN with the RCS average temperature > 200*F but < 350*F for a period not to exceed 7 days. If the inoperable PORV is not restored to OPERABLE status within 7 days, then completely depressurize the RCS and vent through one open PORV or an equivalent size opening within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, b
One PORV may be inoperable in COLD SHUTDOWN or REFUELING SHUTDOWN with the reactor vessel head bolted for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inoperable PORV is not restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> then completely depressurize the RCS and vent through one open PORV or an equivalent size opening within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Amendment Nos. 207 and 207
L h
s TABLE 3.1-1 i
2 i
UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d) t l
NMWD(b)
}
UPPER SHELF HEAT OR MATERIAL.
Cu M
TNDT RTNDT ENERGY MATERIAL CODE NO.
SPEC.NO.
g 3
m g
Closure head dome C4315-2 A53313 Cl.1
.14
.59 0
0 75 Head flange FV-1894 A508 Cl. 2
.13
.64 10(a) 10 125 I
Vessel flenge FV-1870 A508 Cl. 2
.10
.65 10(a) io 74 l
l Inlet pa nia 9-5078 A508 Cl. 2
.87 60(a) 60 64 Inlet nozzle 9-4819 A508 Cl. 2
.84 60(a) 60 68 F
inletpania 9-4787 A508 Cl. 2
.85 60(a) 60 64 Outlet nozzle 9-4762 A508 Cl. 2
.83 60(a) 60 85
{
.04 60(a) 60 72 Outlet nozzle 9-4788 A508 Cl. 2 Outlet rmnia 9-4825 A508 Cl. 2
.85 60(a) 60 68
[
i I
Upper shot 122V109 A508 Cl. 2
.09
.74 40 40 83 3
C4326-1 A533B Cl.1
.11
.55 10 10 115(c) k Intermediate shell a
i intermodate shou C4326-2 A5338 Cl.1
.11
.55 0
0 94 tmuer shell C4415-1 A5338 Cl.1
.11
.50 20 20 103(c) l Lower shou C4415-2 A5338 Cl.1
.11
.50 0
0 83 y
m i m
S ScGom head rind 123T338 A508 Cl. 2
.69 50 50 86 i
Bonom dome C4315-3 A5338 Cl.1
.14
.59 0
0 85 A,
a
~
O Inter. & lower shou 8T1554 & Linde 80 flux
.18
.63 0(a)
-5 77/EMA(e) verticalweed seam i
L1, L3, & L4 f
i
I e
e a
l TABLE 3.1-1 (Continued)
[
UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d) i NMWD(b)
UPPER SHELF HEAT OR MATERIAL Cu M
TNDT RTNDT ENERGY MATERIAL CODE NO.
SPEC.NO-N N
M M
IFTLE j
Lower shellvertical 209L44 & Linde 80 flux
.35
.68 0(a)
-7 70/EMA(e) weld seam. L2 inter.to lower 72445 & Linde 80 flux
.21
.59 0(a)
-5 77(a)/EMA(e) shell girth seam i
Upper shellto inter.
25017 & SAF 89 flux
.33
.10 0(a)
O EMA(e) shou gisth seam i
NOTES:
?
E
$ (4 Estimated per NRC standard review plan NUREG-0800. Sechon MTEB 5-2 o.s (b) Normal to major wodung direden - eshmated per NRC standasd review plan NUREG-0800. Section MTEB 5-2 g (c) Adualvalues v.
(d) Reactor Vessel Fahncator Certified Test Reports
~
(e) The approved equivalent marges analysas in the Topical Reports BAW-2192PA and BAW-2178PA demonstrates ces ' _- s with the requrements of d i
g 10 CFR 50. Appendix G.
g 7
~
m N
l
TABLE 3.1-2 UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) i fe4WD(b) f UPPER SHELF HEAT OR MATERIAL Cu Ni TNDT RTNDT ENERGY l
MATERIAL CODE NO.
SPEC.NO.
S S
S S
(FTLB1 Closure head dome C4381-2 A5338 Cl.1
.15
.52
-20 7
81 l
Head flange ZV-3475 A508 Cl. 2
.11
.60
<10(a)
<10 129 Vessel flange ZV-3476 A508 Cl. 2
.10
.64
-65(a)
-65 129 L
.87 60(a) 60 66 Inlet nozzle 9-4815 A508 Cl. 2 i
Inlet nozzle 9-5104 A508 C1. 2
.84 60(a) 60 73 i
inlet nozzle 9-5205 A508 Cl. 2
.86 60(a) 60 66 l
.85 60(a) 60 74 Outlet nozzle 9-4825 A508 Cl. 2 Outlet nozzle 9-5086 A508 Cl. 2
.86 80(a) 60 79 l
Outlet nozzle 9-5086 A508 Cl. 2
.87 60(a) 60 73 I
l' Upper shes 123V303 A508 Cl. 2
.09
.73 30 30 104 intermediate shou C4331-2 A5338 Cl.1
.12
.60
-10
-10 84 2-3 Intermediate shou C4339-2 A5338 Cl.1
.11
.54
-20
-20 83 o
k Lowershes C4208-2 A5338 Cl.1
.15
.55
-30
-30 94
}
s t
[ Lowershou C4339-1 A5338 Cl.1
.11
.54
-10
-10 105(c) l Bottom head ring 123T321 A508 Cl. 2
.71 10 10 101 m
" i
$ Bottom dome C4361-3 A5338 Cl.1
.15
.52
-20
-15 80 y:
i intermediate shou 72445 & Linde 80 tium
.21
.59
-5 77(a)/EMA(d) verticalweld seams Lot 8579 y
L3 (100%). L4 (OD50%)
o L4 (ID50%)
8T1762 & Linde 80 tiux 8597
.20
.55
-5 EMA(d)
l l
TABLE 3.1-2 (Contmuod)
UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) l i
NMWD(b)
UPPER SHELF HEATOR MATERIAL Cu M
TNDT RTNOT ENERGY MATERIAL CODE NO.
SPEC.NO.
CB On M
M
[FT LB)
{
Lower shou vencalweeds Soam L2(ID 63%)
ST1762 & Linde 80 flux 8507
.20
.55
-5 EMA(d) l Seam L1 (100%)
8T1762 & Linde 80 flux 8507
.20
.55
-5 EMA(d) i Seam L2 (O037%)
8T17s2 & Linde 80 flux 8632
.20
.55
-5 EMA(d) i inter. lo lower 0227 and Grau La Flux LW320
.19
.56 0(a) 0 90(c)mgA(d) l shellginh seam e
Upper shell to inter.
4275 & SAF 89 Ilux
.35
.10 0(a) 0 EMA(d)
{
i shallginh seam i
NOTES:
i i
2 (a) Estimated per NRC standard review plan, NUREG-0800, Secton MTEB 5-2 (D) Normal to major wodung dredson - eshmeted per NRC standasd review plan, NUREG-0000 Sedion MTEB 5-2 s
(c) Adual value based on sunetance tests normal to the mapor wondnD drection 5
(d) The appmved equwalent margms analyses in the Topical Reports BAW-2192PA and BAW-2178PA demonstrates compliance wth the requemments of 10 CFR 50, Appendix G.
m 5
L b
N bl l
i
Figure 3.1-1 Surry Units 1 and 2 Reactor Coolant System Heatup Limitations Meterial Property Base Limiting Meterial Sorry Unit 1 Intermedete to Lower Shen Cire Wold Umiting Adpusted RT(NOT)(Surry 1 et 28.8 EFPY):
228.4F (1/4.T),189.5 F (3/4 T) 2500.00 t
1 3
I l
' l : Leek Test Lirmt - f
/
^ 2000.00
/'
g i.
/
t
[
l iji'
.I
/.ll 3
ii US
,'i, t
g,,,.'
I l..'
i li y
n.
i i
!it i
,i 4
- y i
i: i i
e
/
m
- UrW=Na l['
l l
I g
operaten
/-
e
,1 1000.00
/'
i,'
lii:
iii
- iii, gg g
i-ii.
'l.!
.p 4 i4 e j
ii.:
li ii, g
- yf i
operaoa i!.,
,e s
Rmesi! ;
i y
(FMr) se j,
i
.E 500.00 ao p-i 7
p so t
i'.
1, i
ll: l l
0.00 l
1 0
50 100 150 200
,250 300 350 400 Indicated Cold Leg Temperature (Deg. F) l Surry Unas 1 and 2 Reactor Cooient System Hootup Limitatione (Heste Rates up to 6o F/hr) Aem for the First 28.8 EFPY for Surry Und 1 and the First 29.4 EFPY for Surry Unit 2 Amendment Nos. 207 and 207 9
Figure 3.1-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations Matenal Property Basis LMt9; *1sterial: Surry Unit 1 Intermediate to Lower Shen Circ Wald Lirruting Adpuoted RT(NDT) (Surry 1 at 28.8 EFPY):
228.4 F11/4 T),189.5 F (3/4 T) 2500.00 i
li' I
i i
i
,i;;
i
/
i i
l
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