ML20058C430
| ML20058C430 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 10/18/1990 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058C434 | List: |
| References | |
| GL-88-11, NUDOCS 9011010240 | |
| Download: ML20058C430 (12) | |
Text
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[OS "ICu,)g UNITED STATES
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NUCLEAR REGULATORY COMMISSION-s# : gE g, 3 WASHINGTON. D C. 20$55 k..v )
ILLIN0IS POWER COMPANY, ET AL.
DOCKET NO. 50-461 CLINTON POWER STATION, UNIT N0.1 -
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.51 L.icense No. NPF-62 1.
The Nuclear Regulatory Comission-(the Comission) has found that:
i A.
The application for amendment by Illinois Power Company * (IP),
and Soyland Power Cooperative, Inc. (the licensees) dated April 27, 1990 complies with the standards and requirements of the Atomic Energy Act of 1954,- as amended (the Act), and the Connission'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; l
C.
There is reasonable assurance (i)lthat.the activities authorized by l
1 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 E.
The issuance of this emendment is in:accordance with 10 CFR Part 51 i
of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly the license-is amended by changes-to the Technical Specifi-cationsasIndicatedintheattachmenttothislicenseamendment,and paragraph 2.C.(2), of Facility Operating License.No. NPF-62 is hereby amended to read as follows:
d..
FT'llinals Power Company is authorized to act as agent for Soyland Power Cooperative, Inc. and has exclusive responsibility and control over the physical construction, operation andl maintenance of'the facility.
9011010240 901019 s
DR ADOCK 0500o461 PDC
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(2) Technical Specifications ana Ervironmental Protection Plan The Technical Specifications contained in Appendix A and the-Environmental Protection Plan contained in Appendix B, as revised through Amendment No. SI
, are hereby incorporated into this license.
Illinois Power Company chall 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.
FOR THE NUCLEAR REGULATORY COMMISSION i
John N. Hannon, Director Project Directorate III-3 Division'of. Reactor Projects - III, IV, V and Special Projects-Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of Issuance: October 18, 1990 i
i l
~~.
j ATTACHMENT TO LICENSE AMENDMENT NO. 51 FACILITY OPERATING LICENSE NO. NPF '
DOCKET NO. 50-461' Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Insert xviii xviii 3/4 4-22 3/4 4-22 3/4 4-23
-3/4 4-23 3/4 4-24 3/4 4-24 3/4 4-25 3/4 4-25 B 3/4 4-6 B 3/4 4-6 B 3/4 4-8 8 3/4 4-8 B 3/4 4-10 B 3/4 4-10 t
B 3/4 4-11 B'3/4'4-11 l
i v"
~k
i
.INDEX-l
-l I
BASES l
l SECTION PAGE INSTRUMENTATION (Continued)
Main Condenser Offgas Treatment System Explosive
(
Gas Monitoring Instrumentation............................
B 3/4 3-8 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM.......................
B 3/4 3-8 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION...................
B 3/4 3 3/4 3.10 NUCLEAR SYSTEM PROTECTION SYSTEM - SELF TEST SYSTEM.......
B 3/4 3-9 Bases figure B 3/4.3-1 Reactor Vessel Water Leve1..................
B 3/4 3-10 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 4-3 3/4.4.3 REACTOR COOLANT 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 1
3/4.4.5 SPECIFIC ACTIVITY.........................................
B 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS...............................
B 3/4 4-5 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES..........................
B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY......................................
B 3/4 4-7 3/4.4.9 RESIDUAL HEAT REM 0 VAL...................................
B 3/4 4-7 Bases Table B 3/4.4.6-1 Reactor Vessel Toughness Values............
B 3/4 4-8 Bases Figure 'B 3/4.4.6-1 Fast Neutron Fluenc? (E>l Mev) at I.D.
[
surface as a Function of Service Life.....................
B 3/4 4-10 Figure B 3/4.4.6-2 DELETED.........................................
B 3/4 4-11 CLINTON - UNIT 1 xviii Amendment No. 51
_ r REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION l
3.4.6.1 The reactor vessel pressure and metal temperature shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydro-static or leak testing; (2) curve B for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for i
operations with a critical core other than low power PHYSICS TESTS, with:
a.
The maximum rate of change of reactor vessel steam space coolant tempera-ture during normal heatup or cooldown shall be limited to 100'F in any I hour, b.
A maximum metal temperature change of < 20 F in any I hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and c.
The reactor vessel flange and head flange metal temperature shall be
> 70'F when reactor vessel head bolting studs are under full tension.
APPLICABILITY:
At all times, i
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-ofvlimit 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 vessel pressure and metal temperature of the-reactor vessel flange surfaces, bottom head outside surface and bottom head inside surface, as measured by the bottom head drain temperature, shall be determined to be within the operating limits defined by Figure 3.4.6.1-1 at least once per 30 minutes.
1 4.4.6.1.2 The reactor steam space coolant temperature shall be determined to j_
be within the heatup and cooldown limits of 100 F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at least once N
per 30 minutes, i
l CLINTON - UNIT 1 3/4 4-22 Amendment No. 51
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REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.3 The reactor coolant system temperature and pressure shall be detcc-mined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curve within 15 minutes prior to the withdrawal of control rods to bring the l
reactor to criticality.
4.4.6.1.4 The reactor vessel material specimens shall be removed and examined to determine changes in reactor pressure vessel material properties as a l
function of time and THERMAL POWER as required by 10 CFR 50, Appendix H, in-accordance with the schedule in Table 4.4.6.1-1.
The results of these examinations shall be used to adjust the curves of Figure 3.4.6.1-1.
l 4.4.6.1.5 DELETED.
4.4.6.1.6 The reactor vessel flange and head flange temperature shall be verified to be > 70'F when vessel head bolting studs are under full tension:
a.
In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
1.
< 90'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.
580F,atleastonceper30 minutes, b.
Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs except 10 percent of the bolting studs may be fully tensioned at 3 10 F but 1 70 F.
l l
A l
CLINTON - UNIT 1 3/4 4-23 Amendment No. 51
1600 CURVE A EFPY g '#
9 C
1400
'i I
1200 b
h i
1000 N
m 800 h
8 AND C - CORE BELTUNE WITH ASSUMED 130*F g
SHIFT FROM AN INITIAL l
WELD RTNOT OF - 30'F '
CURVES A. 8 AND C ARE VAUD 600 FOR 12 EFPY OF OPERATION f
CURVE A INCLUDES BELTUNE ART VALUES SHOWN BELOW,-
t l
58 8
88-
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12 100 312 psig A - SYSTEM HYDROTEST UMIT WITH FUEL IN VESSEL 8 - NON. NUCLEAR HEATING 200
. UMIT l
BOLTUP C - NUCLEAR (CORE CRITICAU l~
70'F UMIT I'
I I
I I
0 0
100 200 300 400 500
.600 y..
MIMMUM REACTOR VESSEL METAL TEMPERATURE ('F)
Figure 3.4.6.1-1. Reactor Vessel Pressure Versus Minimum Reactor Vessel Metal Temperature l
- CLINTON - UNIT 1-3/4 4-24 Amendment NO.'51
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E TABLE 4.4.6.1-1
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n REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE E
CAPSULE VESSEL LEAD WITHDRAWAL TIME E
NUMBER LOCATION FACTOR at I.D.
(EFPY)
--e 1.
Capsule 1 3
0.67 10 2.
Capsule 2 177 0.67 20 3.
Capsule 3 183 0.67 Spare R
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REACTOR COOLANT SYSTEM' BASES 3/4.4.6 PRESSURE / TEMPERATURE LIMITS (Continued)
The reactor vessel materials have been tested to determine their initial RTNDT' The results of these tests are shown in Table B 3/4.4.6-1.
Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT f the core beltline region.
Therefore, an adjusted NDT reference temperature, based upon the fluence, nickel content and copper con-tent of the material in question, can be predicted using Regulatory Guide 1.99,
" Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.
The pressure / temperature limit curve, Figure 3.4.6.1-1, curves A, B, and C, includes an assumed shift in RT f r the conditions at 12 Effective Full NDT Power Years.
The actual shif t in RT of the vessel material will be estab-NDT lished periodically during operation by removing and evaluating, in accordance with ASTM E185 and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.
The irradiated specimens can be used to predict reactor vessel material transition temperature shift.
Flux wires which were removed after the first fuel cycle and will be removed at later intervals with the surveillance speci-mens are analyzed and provide an improved neutron fluence estimate for the reactor vessel.
This data is then used to modify Bases Figure B 3/4.4.6-1 and predictions of reactor vessel material transition temperature shift per Regu-latory Guide 1.99, Revision 2. The operating limit curves of Figure 3.4.6.1-1 have been and will be adjusted, as required, on the basis of the specimen data and the recommendations of Regulatory Guide 1.99, Revision 2.
The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C'and A l
for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with tha minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.
The number of reactor vessel irradiation surveillance capsules and the frequen-cies for removing and testing the specimens in.these capsules are provided in Table 4.4.6.1-1 to assure compliance with the requirements of Appendix H to 10 CFR 50.
3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each sf the main steam lines to minimize the potentiaPleakage-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 considerations 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.
CLINTON - UNIT 1 B 3/4 4-6 Amendment No. 51
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n BASES TABLE B 3/4.4.6-1 C$
REACTOR VESSEL TOUGHNESS VALUES e.
LIMITING HEAT #-SLAB #
MIN. EOL E
BELTLINE WELD SEAM I.D.
OR STARTING MAX. EOL NDT ( F)* UPPER SHELF U
I.
COMPONENT OR MAT'L TYPE HEAT #/ LOT #
CU(%) Ni(%)
(FT-LBS) w PLATE SA-533 GR.8,
-C 4380-2 0.07 0.063 -20 69 85 49 CL1 WELD N/A 76492/
0.10 1.08
-30 164 76 134 L430827AE NOTE:
- These values are given only for the benefit of calculating the end-of-life-(EOL)(32EFPY)
co R
HEAT # - SLAB #
HEAT # - SLAB #
NON-BELTLINE MT' L TYPE OR OR HIGHEST-STARTING
[-
RTNOT ( F)
I.
COMPONENT WELD STEAM I.D.
HEAT #/ LOT #
SHELL RING SA-533 GR.B CL.1 C4240-2
- -10 A2758-1 BOTTOM HEAD DOME.
-A2757-1
-10 BOTTOM HEAD TORUS C4027-1
+10 TOP HEAD DOME-C4374 -40 TOP HEAD TORUS
'A2879-2
-10 5.
TOP HEAD FLANGE SA-508, CL. 2 CCZ 41-5478
-40 5
'SER 915 z
P SER 878-
.S' FEEDWATER N0ZZLE Q2AL10W
-20
._,Y
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~-
_y
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f 10 I
i 9
l 1
e 1
7 NOMINAL l.D.
?2 5
e
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E 5
3 6
w W
5 c'
g4 B
W 3
i I,
l.
I 2
1 I
I I
I O
O 10 20 30 40 50 SERVICE uFE (yeets')
Bases Figure B 3/4.4.6-1. Fast Neutron Fluence (E>1 MeV) j at I.D. Surface as a Function of Service Life
'At 90% Rated Thermal Power and 90% Availability l
CLINTON - UNIT 1 l
83/44 Amendment No. 51
1 i
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Bases Figure B 3/4.4.6-2 DELETED CLINTON - UNIT 1 B~3/4 4-11 Amendment No. E'
.