ML20126F840
| ML20126F840 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/18/1992 |
| From: | Hall J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126F844 | List: |
| References | |
| NUDOCS 9212310169 | |
| Download: ML20126F840 (16) | |
Text
.-
O'
.S y'
n UNITED STATES >
)
^ :!
si NUCLEAR REGULATORY COMMISSION l
f wAssiNotoN. D.c. aohE6
)
%...../
i IliE CLEVELAND ELECTRIC ILLUMINATING COMPANY. ET AL.
DOCKET NO. 50-440 PERRY NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 45-License No. NPF-58 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by The Cleveland Electric illuminating Company, Centerior Service Company, Duquesne Light.
Company, Ohio Edison Com)any, Pennsylvania Power Company,-and-'
Toledo Edison Company (tle licensees) dated September 14, 1990, revised and sup)1emented March 15, 1991-and December 20, 1991,
-complies with tie standards and requirements of the Atomic Energy.
Act of 1954, as amended (the Act), and the Commission'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 Commission; that the activities' authorized There is reasonable assurance'(i)d without endangering the health C.
by this amendment can be conducte and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission'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 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.
2.
Accordingly, the license is amended by changes to the-Technica1JSpecifi -
cations as-indicated-in the~ attachment to this license amendment, and paragraph 2.C.(2)of'FacilityOperatingLicenseNo.NPF-58is~hereby amended to read as follows:
9212310169 921215 fDR ADOCK 05000440
. PDR,
i.
(2)
Itchnical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix 8, as revised through Amendment No. 45 are hereby incorporated into this license.
The Cleveland Electric illuminating Company shall operate the facility in accordance with the Technical Specifi-cations and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 11$rt.0A Jame
. Hall, Sr. Project Manager Project Directorate III-3 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuanca: December 13, 1992 I
r ATTACHMENT TO LICENSE AMENDMENT NO. 45 FACILITY OPERAllNG LICENSE NO. NPF-58 DOCKET NO. 50-440 Replace the following pages of the Appendix 'A' Technical Specifications with the attach 6d pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.
Remove Insert xi xi XX XX 3/4 4-20 3/4 4-20 3/4 4-21 3/4 4-21 3/4 4-22 B 3/4 4-5 8 3/4 4-5 B 3/4 4-5a B 3/4 4-6 B 3/4 4-6 8 3/4 4-7 B 3/4 4-7 B 3/4 4-8
- 4 4-8
i LIMITING CONVITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAG,E 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.....................
3/4 4.......
Operational Leakage.....................................
3/4 4-10 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves..................
3/4 4-12 3/4.4.4 CHEMISTRY...............................................
3/4 4-13 Table 3.4.4-1 Re6ctor Coolant System Chemistry Limits..............................
3/4 4-15 3/4.4.5 SPECIFIC ACTIVITY.......................................
3/4 4-16 Table 4.4.5-1 Primary Coolant Specific Activity Sample end Analysi s Program.........
3/4 4-18 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...........................
3/4 4-19 Figure 3.4.6.1-1 Reactor Vessel Pressure Versus lietal Temperature Valid Up to 8 EFPY 3/4 4-21 Table 4.4.6.1.3-1 Deleted R e a c to r S te am D om e......................................
3/4 4-23 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................
3/4 4-24 3/4.4.8 STRUCTURAL INTEGRITY....................................
3/4 4-25 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown............................................
3/4 4-26 Cold Shutdown...........................................
3/4 4-27 3/4.5 EMERGENCY CORE COOLING SYF'"U.S 3/4.5.1 ECCS - 0PERATING........................................
3/4 5-1 PERRY - UNIT 1 xi Amendment No. 45
)
l l
s LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE EMERGENCY CORE COOLING SYSTEM (Continued) 3/4.5.2 ECCS - SHUTD0WN.........................................
3/4 5-6 3/4.5.3 SUPPRESSION P00L........................................
3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity - Operating...............
3/4 6-1 Primary Containment Integrity - Shutdown................
3/4 6-2 Primary Containment Leakage.............................
3/4 6-3 Primary Containment Ai r Loc ks...........................
3/4 6-6 MSIV Leakage Control System.............................
3/4 6-8 Containmen+. Structural Integrity........................
3/4 6-9 Containment Internal Pressure...........................
3/4 6-10 Primary Containment Average Ai r Temperature.............
3/4 6-11 Drywell and Containment Purge System....................
3/4 6-12 Feedwater Leakage Control System........................
3/4 6-14 3/4.6.2 DRYWELL D rywe l l I nte g r i ty.......................................
3/4 6-15 Orywell Bypass Leakage..................................
3/4 6-16 Drywell Air Lock........................................
3/4 6-17 O rywel l ' Structural Integri ty............................
3/4 6-19 Orywell Inte rnal Pre s sure...............................
3/4 6-20 Drywell Average Air Temperature.........................
3/4 6-21 3/4.6.3 DEPRESSURIZATION SYSTEMS Suppression Poo1........................................
3/4 6-22 PERRY - UNIT 1 xii
LBASES SECTION PAGE REACTOR CDOLANT SYSTEM (Continued).
3/4.4.5 SPECIFIC ACTIVITY.......................................-
B 3/4l4-4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.............................-
B 3/4 4-5 Bases Table B 3/4 4.6-1 Reactor Vessel Toughness.................
8-3/4 4-7 Bases Figure B 3/4 4.6-1 Fast Neutron Fluence (E>l MeV) At-Inside Surface l
As A Function of Service 1
Life.....................-
B 3/4 4-8 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................-
B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY....................................-
- B-3/4 4-6 3/4.4.9 RESIDUAL HEAT REM 0 VAL...................................
- B 3/4 4-6 3/4. 5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN.................
B'3/4 5-1 3/4.5.3 SUPPRESSION P00L........................................
B 3/4'5-2
-3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Prima ry Containment Integri ty...........................
B 3/4'6-1 P rima ry Contai nment Lea kage.............................
Bi3/4 6-1 Containment AirlLocks...................................
B 3/4'6-2?
MSIV Leakage Control System.............................
B 3/4 6-2s Containment Structural Integrity.......................
_B 3/4 6-2:
g Contai nment Inte rnal ~ Pres s ure........................... --
B-3/4 6-2a;
~
Contai nment Ave rage Ai r Temperature.....................-
B-3/4 6 2a Drywell and Containment Purge System....................
B 3/4 6-2a-Feedwater Leakage _ Control System.........................
B 3/4 6 ~ PERRY - UNIT 1 xx knendment tur. l$. 45:
-*ww vv m -v - t -vv
-v---
w
-y
>-e--
+w r
wmv et-sr
--+e
--,-.-w
- m. c - ww
-ry, yevr-
- wr-,-,,we
1'
+
s BASES SECTION PAGE
-INSTRUMENTATION (Continued) 3/4.3.7 MONITORING INSTRUMENTATION-Radiation Monitoring Instrumentation....................
B 3/4 3 Seismic Monitoring Instrumentation......................
B 3/4 3-4 Meteorological Monitoring Instrumentation...............
B 3/4 3-4 Remote Shutdown System Instrumentation and Controls.....
B 3/4 3-5 Accident Monitoring Instrumentation.....................
B 3/4 3-5 Source Range Monitors....................................
B 3/4 3-5 Traversing In-Core Probe.Systes.........................-
B-3/4 3-5--
Loose-Part Detection System.............................
B 3/4 3-6 Radioactive Liquid-Effluent Monitoring Instrumentation.........................................
B 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instrumentation.........................................
- B 3/4 3-6 3/4.3.8 TURBINE OVERSPEED' PROTECTION SYSTEM.....................
B 3/4 3-6 3/4'. 3. 9 PLANT SYSTEMS ACTUATION INSTRUMENTATION.................
B 3/4.3-6 Bases Figure B 3/4 3-1 Reactor Vessel Water Leve1......................
B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM....................................
B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES....................................
B 3/4 4-2 3/4.4.3 REACTOR'C00LANT SYSTEM LEAKAGE Leakage. Detection Systems...............................-
B 3/4 4-3 --
Operational Leakage.....................................
B 3/4 4-3 3/4.4.4 CHEMISTRY...............................................
B 3/4-4-4 PERRY - UNIT 1 xix l
4 m-m__m
__m-__.
1 REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1: _ (1) curves A and A' for hydrostatic or leak testing; (2) curves 8 and B' for heatup by non-nuclear!
means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C and C' for operations with a critical core other than low power -
PHYSICS TESTS, with:
a.
A maximum heatup of 100*F in any one hour period, b.
A maximum cooldown of 100'F in any one hour period, c.
A maximum temperature change of less than or equal to 20'F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.
The reactor vessel flange and head flange temperature greater than or equal to 70*F when reactor vessel head-bolting studs are.under.
tension.
APPLICABILITY:
At all times.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limits 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 operations or be in at least HOT _ SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within-the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup,.cooldown and inservice leak and hydrostatic -
testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to-the right of the-limit lines,of Figure 3.4.6.1-1 curves A and A',
8 and 8', or C and C', as applicable, at least once per 30 minutes.
1 PERRY - UNIT 1 3/4 4-19
s REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature end pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curves C and C' within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and at least once per 30 minutes during system heatup.
4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed and examined to determine changes in reactor pressure vessel material properties as required by 10 CFR 50, Appendix H.
The results of these examinations stall be used to update the curves of Figure 3.4.6.1-1.
4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be areater than or equal to 70'F:
a.
In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
1.
5,100*F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
5,80'F, at least once per 30 minutes.
b.
Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.
PERRY - UNIT 1 3/4 4-20 Amendment No. 45
o taso -....... 3........
........:. I...
>...... t.. n n o.;. n. n n y n. n o t.. o. n. p. n n o y. n n o t.
l.
l
.(.......:........l.......<........;...............(.
......4 c g c..
s g%
l troo AA.
.o.
.......t........ 7.....
.;......t.
.......y.............. t.. n n n 9. n. n o g o n. n t.o. n o.
l i
m
.a.,.
1000 - -... " i. " ". n t. o n "l -
~"
on"i..."."." " "i.n." ni." " '.; n " " 4." " " i.
.g
.g c
,.r e
.,w f
f
[]
eco
..y.
.y.......;....... p.......y.......g........;....... 3.......;
g w
c 2
AC es*
cC'.
- Goo --
....n...u...
=. o u
.=.=n..n.'n'...n..'.= n.u n u.
n. n n... n u n.
n.'.
-n
,t, I
I
.J P
A - SYSTEM HYDRO OR LEAK TEST
[]
8 R
- R LIMIT WITH FUEL IN VESSEL
-l
. -l w
g 4eo -
4.....
g wog. NUCLEAR HEATUP/COOLDOWN I
l AND PHYSICS TEST LIMIT C - NUCLEAR (CORE CRITICAL) LIMIT 312.5 PSIG i
A*.B*.C*=A.B.C LIMITS AFTER 4 EFPY A.A.
e.e '
cc 200
.......t,... <....:-.....!.......t......
LIMITS DURING FIRST 4 EFPY f
= = = LIMITS BETWEEN 4 C B EFPY
. nottug 7o*F
- o i
I I
I I
I I
I I
I I
o too too soo 400 soo MINIMUM REACTOR VESSEL HETAL TEMPERATURE (*F)
REACTOR VESSEL PRESSURE VERSUS HETAL TEMPERATURE. VALID UP TG B EFPY Figur e 3. 4.6.1-1 s
PERRY - UNIT i 3/4 4-21 Amendment No. 45
'l
^
L r
u l
This page intentionally left blank.
I PERRY - UNIT 1 3/4 4-22 Amendment No. 45
_____2
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Contined)
Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.
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.
3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the ef fects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section 3.9 of the USAR.
During startup l
and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The purpose of this specification is to establish operating limits that provide a wide margin to brittle failure for major piping and pressure vessel components of the Reactor Coolant Pressure Boundary (RCPB).
RCPB materials are subject to brittle failure below their nil-ductility temperature (NDT), at relatively low stresses.
Below NDT, stresses are carefully limited by specifying both allowable pressure and heatup/cooldown rates. Of the major components within the RCPB, the reactor vessel is the component most subject to brittle failure and therefore is the component for which these technical specification limits are most pertinent.
The basis of the pressure and temperature (P-T) limits is found in Appendix G to 10 CFR 50. Appendix G requires that the limits be based on specific fracture toughness requirements for RCPB materials such that an adequate margin to brittle failure will be provided during operational t
occurrences.
10 CFR Part 50 Appendix G mandates the use of ASME Section III.
Appendix G.
The concern addressed by Appendix G is that undetected flaws could exist in the RCPB components, which under certain reactor coolant system P-T combinations could cause stress concentrations at flaw locations resulting in flaw growth to failure before the ultimate strength of the material is attained.
Flaw growth is resisted by the material toughness, a property that increases with temperature.
Furthennore, the material toughness is affected by neutron fluence which causes the steel ductility to decrease. The effect of fluence is cumulative, and ductility steadily decreases with exposure time. Toughness is also dependent on the chemistry of the base metal and its impurities.
Table B 3/4 4.6-1 provides initial and predicted end-of-life PERRY - UNIT 1 B 3/4 4-5 Amendment No. 45
)
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) reactor vessel toughness data.
Fracture toughness limits and the basis for compliance are more fully discussed in Chapter 5 of the USAR.
Although any region within the pressure boundary is subject to non-ductile failure, the regions that provide the most restrictive limits are the vessel closure head flange, the feedwater nozzles, the control rod drive nozzles, and the vessel beltline. At any specific pressure, temperature, and temperature rate of change, one location witnir. the geometry of the reactor vessel will dictate the most restrictive limit. Across the entire pressure /
temperature span of the limit curves, the curves are composites of the most restrictive regions.
The reactor vessel beltline is the only RCPB material which experiences significant neutron fluence. Since fluence causes an increase in NDT, the vessel beltline becomes most limiting (requires the highest temperature when pressurized) af ter 4 years of operation.
The value of NDT used to set operating limits is called the nil-ductility reference temperature (RTNDT) which incorporates safety margin for variation in material properties and measurement.
The actual shift in RT DT of the vessel material will be established N
periodically during operation by removing and evaluating irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area, in accordance with ASTM E185 and 10 CFR 50, Appendix H.
The operating limit curves (A, B, C) are adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99 Revision 2
" Radiation Embrittlement of Reactor Vessel Materials," to assure a ductile pressure boundary.
Operating limits for pressure and temperature are provided for three categories of operation:
(a) hydrostatic pressure tests and leak tests, referred to as Figura 3.4.6.1-1 Curve A; (b) non-nuclear heatup/cooldown and low-power Physics Tests, referred to as Curve B; and (c) core critical operation other than during Physics Tests, referred to as Curve C.
The beltline region minimum temperatere limits are adjusted to account for vessel irradiation in curves A', B', C'.
These curves have been developed for a large number of operating cycles and provide a conservative margin to non-ductile failure. Although they have been created to provide limits for normal operations, they also can be used as a basis for determining if evaluations are necessary for abnormal transients which can begin from power operation. ASME Section XI Appendix E provides a recommended methodology for evaluating operating events which cause an excursion outside of the normal limits.
PERRY - UNIT 1 B 3/4 4-Sa Amendment No.45 I
' /
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
A sensitivity study was performed for a BWR Class 3 plant to investigate the effects of increasing the initial reactor pressure relative to the initial value used in the overpressure protection analysis on the peak system pressure.
This analysis showed that increasing the initial operating pressure results in an increase in peak system pressure that is less than half the initial pressure increase.
For Perry, the Technical Specification limit on the high-reactor-pressure scram is 1061.7 psig.
Therefore, since the vessel dome pressure used in the overpressurization analysis was 1045 psig, the maximum increase in the initial pressure would be limited to 50 psi, and the maximum peak system pres-sure increase during the overpressure design transient would be less than 25 psi.
Recirculation pump trip has resulted in an increase of 2 to 6 psi in calcula-tions for other BWRs.
These results indicate that considerable margin is avail-able to Perry before reaching the Code limit and that GDC 15 will be satisfied even if increased initial dome pressure and recirculation pump trip are con-sidered.
Since the Perry specific overpressure analysis (as well as all other transient analyses in Chapter 15 of the FSAR) were performed assuming an ini-tial dome pressure of 1045 psig, the Technical Specifications operating pres-sure limit is 1045 psig.
3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.
Only one valve in each line is required to maintain the integrity of the containment, however single failure con:;iderations require that two valves be OPERABLE.
The surveillance requirements are based on the operating history of this type valve.
The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.
The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.
3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of thase components will be maintained at an acceptable level throughout the life of the plant.
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 1977 Edition and Addenda through Summer, 1978.
The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).
3/4.4.9 RESIOUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capa-bility for removing core decay heat and mixing to assure accurate temperature indication; however, single failure considerations require that two loops be i
OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.
PERRY - UNIT 1-B 3/4 4-6
~., _ -
BASES TABLE B 3/4 4.6-l' M;
REACTOR YESSEL TOUGHNESS HIGHEST I.
BELTLINE gyggggg STARTIf). UPPER SHELF.
32 E R RTNOT( F ENERGY (Ft-Lb)
M COMPONENT MATERIAL-HEAT #/ LOT #
CU(%)
Ni(%)
.. RTNOT(,F) j
-Plate SA533 GrB C2557-1
.06 0.61
+10 75.
+60..
Class 1 4
Weld BD,BF 627260/8322A27AE'
.06 1.08
-30 88'
+80 II.
NON-BELTLINE HIGHEST STARTING l'
COMPONENT" MATERIAL RTNOT( F)
Shell Ring SA 533 Gr.B. C1.1
+10 Botton Head SA 533 Gr.B, C1.1
+10 e
n Top Head SA 533 Gr.B. C1.1
+10
[
+10.
. 10.
+
}
Feedwater Nozzle ;
SA 508,.C1.2
-20:
- )
Weld' Low alloy steel'per i
g GE purchase specification
-.20
~
' f Closure Studs SA 540 Gr.823 45 ft-lb"& 25 mils lat. exp.'
requirement met at.+10(*F)-
.m 4
4 y
,m.,c
6
'o 5
N
/
2 4
I
/
"b
/
/
1 h1
[
e
/
/
/
0 0
10 20 30 40 Servk:s Ufa (Years)
Fcat Neutron Fluence (C>1 Wav) ot Ins'de Surfoce os a Function of Service Ufe*
Bases F we D 3/4 4.6-1 9
- At 907, of RATED TIERWL POWER ond 90% ovoBotiEty PutKT - UNIT 1 5 3/4 4-8 Amendment No. 45
)
-.