ML20063K662
| ML20063K662 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/10/1994 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20063K665 | List: |
| References | |
| NUDOCS 9403020009 | |
| Download: ML20063K662 (15) | |
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UNITED STATES gg NUCLEAR REGULATORY COMMISSION g
9j WASHINGTON. D.C. 2055 MOO 1 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION s
DOCKET NO. 50-482 61ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. NPF-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Wolf Creek Generating Station (the facility) Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated May 27, 1993, complies with the 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, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license 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.
9403020009 940210 PDR ADOCK 05000482 P
4
. l 2.
Accordingly, the license is amended by 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-42 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 71, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 30 days of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ODb W, &
Suzanned. Black, Director Project Directorate IV-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 10, 1994 l
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4.
ATTACHMENT TO LICENSE AMENDMENT NO. 71 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
l The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT VIII VIII 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-36 3/4 4-36 B 3/4 4-7 8 3/4 4-7 B 3/4 4-11 B 3/4 4-11
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F b
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)
TABLE 4.3-8 RADIDACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS DELETED Radioactive Gaseous Effluent Monitoring Instrumentation DELETED E xpl osi ve Ga s Monitori ng Instrumentation.................
3/4 3-58 TABLE 3.3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION.............
3/4 3-59 TABLE 4.3-9 EXPLOSIVE GA5 MONITORING INSTRUMENTATION SURVEI LL ANCE REQUI REMENT S..........................
3/4 3-61 3/4.3.4 TUREINE OVERSPEED PROTECTION............................... 3/4 3-63 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation......
3/4 4-1 3/4 4-2 i
Ect Stan:t,.
Het Srcidc.n 3/4 4-3 3/4 4-5 Cold Shatcc,- - Loops Filled.
Cold Shutdown - Loops Not Filled.....
3/4 4-6 3/4.4.2 SArETY VALVE 5 3/4 4-7 Snutdc n 3/4 4-8 Operating.
3/4.4.3 PRE 550RIZER..
3/4 4-9 3/4.4.4 RELIEF VALVE 5...
3/4 4-10 3/4.4.5 STEAM GENERATOR5..........................................
3/4 4-11 TABLE 4.4-1 MINI?iUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION........................
3/4 4-16 TABLE 4.4-2 STE AM GENERATOR TUBE INSPECTION......................
3/4 4-17 3/4.4.6 REACTOR 000LANT SYSTEM LEAKAGE-Leakage Detection 5ystems.................................
3/4 4-18 Operatienal Leakage.......................................
3/4 4-19 WOLF CREEK - UN"T 1 VII Amendment No.15, 42
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...............................................
3/4 4-21 3/4.4.7 CHEMISTRY..............................................
3/4 4-22 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS................
3/4 4-23 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS.........................................
3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY......................................
3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
> 1 pCi/ GRAM DOSE EQUIVALENT I-131..................
3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.....................................
3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...............................
3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 13.6 EFPY..........................
3/4 4-30 l
FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 13.6 EFPY..........................
3/4 4-31 l
TABLE 4.4-5 DELETED Pressurizer..........................................
3/4 4-33 Overpressure Protection Systems......................
3/4 4-34 FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM......................
3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY......................................
3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS..............................
3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULAT0RF.............................................
3/4 5-1 WOLF CREEK - UNIT 1 VIII Amendment No. 40,57,71
3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
i A maximum heatup of 60*F in any 1-hour period for indicated T.,
l a.
less than or equal to 200*F, b.
A maximum heatup of 100*F in any 1-hour period for indicated T,
greater than 200*F, c.
A maximum cooldown of 100*F in any 1-hour period, and d.
A maximum temperature change of less than or equal to 10*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times.
ACTION:
i With any of the'above limits exceeded, restore the temperature and/or pressure 4
to within the limit within 30 minutes; perform an engineering evaluation to l
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 operation or be in at least HOT j
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less i
than200*Fand500psig,respectively,withinthefoWowing30 hours.
j SURV:lllANCE RE0VIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
1 4.4.9.1.2 The reactor vessel material irradiation surveillance. specimens shall be removed and examined, to determine' changes in material properties, as required by 10 CFR Part 50, Appendix H.
The results of these examinations l
i shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4.
j WOLF CREEK - UNIT 1 3/4 4-29 Amendment No. 40,57 j
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MATERIAL PROPERTY BASIS Controlling M terint:
RV Lower Sh ll
~
Copper Cont::nt:
0.07 W ight %
i Nickel Content:
0.62 Weight %
Initial RTuor:
40*F Limiting ART after 13.6 EFPY:
1/4T, 89
- F 3/4T, 79 'F 3,000
.- CRITICALITY LIMIT
~
BASED ON 60'F/HR
. HEATUP CURVE 2,500
-f LEAK TEST LIMIT n
y 2,000 CL CRITICALITY UMIT
@ 1,500 so F/HR BASED ON 100*F/HR c.
HEATUP CURVE HEATUP CURVE O
52
@ 1,000 CRITICALITY LIMIT
~
BASED ON INSERVICE
' ~ ~ ~ " '
[ TEMPERATURE (222*F)
~ ~ -
500 HYDROSTATIC TEST 3g9 pfgg HEATUP CURVE -
FOR THE SERVICE PERIOD UP TO 13.6 EFPY 0
O 100 200 300 400 500 INDICATED TEMPERATURE (DEG. F)
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 13.6 EFPY l
WOLF CREEK - UNIT 1 3/4 4-30 Amendment No. 40,71
[
MATERIAL PROPERTY BASIS Controlling Mitsrial:
RV Lower Sh311 '
^
Copper Content:
0.07 Weight %
Nickel Content:
0.62 Weight %
initial RTuoT:
40*F Limiting ART after 13.6 EFPY:
1/4T,89 'F 3/4T,79 'F 3,000 i._..._4 2,500
.i 9 2.000 cn w
E D
u) cn w 1,500
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w t
k
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COOLDOWN RATES i
5 1,000 g.pgg) g 0
20 40 g-60+
500 -
3 00 __,-
+
.l..._.-_...._-..........__......_.._...._
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100 200 300
-400 500 INDICATED TEMPERATURE (DEG. F).
i FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 13.6 EFPY 5
q WOLF CREEK - UNIT 1 3/4 4-31 Amendment No. 40,71 l
TABLE 4.4-5 i
(Deleted) 1 i
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WOLF CREEK - UNIT 1 3/4 4-32 Amendment No. 57
REACTOR COOLANT SYSTEM SURVEILLANCE REOUTREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
a.
Performance of an AKALOG CHANNEL OPERATIONAL TEST on the PORY actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; b.
Performance of a CHANNE CALIBRATION on the PORY actuatidn channel at least once per 18 months; and c.
Verifying the PORY isolation valve is open at leas't once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:
a.
For RHR suction relief valve 8708B:
By verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RHR RCS Suction Isolation Valves (RRSIVs) 8701B and 8702B are open.
b.
For RHR suction relief valve 8708A:
By verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RRSIVs 8701A and 8702A are open.
c.
Testing pursuant to Specification 4.0.5.
4.4.9.3.3 With the RCS vented, verify the vent pathway at least once per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l WOLF CREEK - UNIT 1 3/4 4-35 Amendment No. M,63
3,000 2,750 1
2,500 TRTD Pswx
( *F)
(PSIG).. :.
f-2,250
- l' 97 485
' ~ ' T.. ~
l 1
32y g
1 2,000 177 485 227 545 i
3 277 745 l
327 1310 l
17)
Eb.1 *750 372 2335
+
E 2 1,500 y
u)
@ 1,250
~
O 1,000 750 500 250
~.
0 100 200 300 400 500 MEASURED RTD TEMPERATURE (DEG. F)
FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM WOLF CREEK - UNIT 1 3/4 4-36 Amendment No. 40,71
9 4
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) b.
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
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 spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583*F.
5.
System preservice hydrotests and in-service 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 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT at the end of 13.6 effective full power years (EFPY) of service life.
Ne,13.6EFPYservicelife period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT of the limEing unirradiated material. The
.mlection of such a limiting k assures that all components in the Reactor Coolant System will be operated c,onservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RT,31; the results of these tests are shown in Table B 3/4.4-1.
Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT,or.
Therefore, an adjusted reference temperature, based upon the fluence and copper content and nickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of-ART ' actor Vessel Materials." computed by Regulatory Guide 1.99, Revision 2, " Radiation Em l
ofEe The heatup and cooldown limit-curves of Figures 3.4-2 and 3.4-3 include predicted' adjustments for this shift in RT,37 at the end of 13.6 EFPY as well as adjustments for possible errors in the pressure l
and temperature sensing instruments.
WOLF CREEK - UNIT I B 3/4 4-7 Amendment No. 40,71 W
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
Values of ART determined in this manner may be used until the results of the next scheduie'd capsule from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H.
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 used to predict the 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 the calculated ART., for.,he equivalent capsule t
radiation exposure.
Allowable pressure-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.
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 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 calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RT 7, is used and this includes the radiation-induced shift, ART 7, 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 heatup or cooldown cannot be greater than the reference stress intensity factor, K,,,
for the metal temperature at that time. X is obtained from the reference fracturetoughnesscurve,definedinAppend,ixGtotheASMECode. The X,,
curve is given by the equation:
K,, - 26.78 + 1.223 exp [0.0145(T-RT., + 160)]
(1) i WOLF CREEK - UNIT 1 B 3/4 4-8 Amendment No. 40,57
10 INNER SURFACE
^
E e
19 E 10 1/4 T 1.88 x 10" 5
5 E
3/4 T 6.69 x 10" w
oz 18
$10 e
z O
k a
W 17 Z 10
~
l 16 10 O
5 10 15 20 25 30 32 35 SERVICE LIFE (EFPY)
FIGURE B 3/4.41 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF EFFECTIVE FULL POWER LIFE i
i WOLF CREEK - UNIT I B 3/4 4-11 Amendment No. 40,71
-=
6 REACTOR COOLANT SYSTEM BASES COOLDOWN (Continued) during cooldown results in a higher value of K at the 1/4T location for IR finite cooldown rates than for steady-state operation.
Furthermore, if conditions exist such that the increase in K exceeds kit, the calculated IR allowable pressure during cooldown will be greater than the steady-state value.
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 f r the 1/4T crack IR during heatup is lower than the K for the 1/4T crack during steady-state yp conditions at the same coolant temperature.
During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K
's f r steady-state and finite heatup rates IR do not offset each other and the pressure-temperature curve based on steady-state 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 of the allowable pressere calculated for steady-state and finite heatup rates is obtained.
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, i
l l
l WOLF CREEK - UNIT 1 B 3/4 4-12 Amendment No. 40 I
.