ML20086C905
| ML20086C905 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/06/1991 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086C908 | List: |
| References | |
| NUDOCS 9111250158 | |
| Download: ML20086C905 (15) | |
Text
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WOLF __ CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50 402 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 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 March 5, 1991, complies with the standards end 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 J.e Commission's regulations and all applicable requirements have been satisfied.
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-Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of facility Operating Licensa No. NPT-42 is bereby amended to read as follows:
2.
Technical Spe,cifications i
The Technical Specifications contained in Appendix A, as revised through Amendesnt No.
51, and the Environmental Protection Plan centained in Appendix d, both of which are attached hereto, are s
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.
FOR THE NUCLEAR REGULATORY COMMISSION S_u C,A~ n. (
Suzann)a,a C. Black, Director Project Directorate IV-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Ragulation
Attachment:
Chtnges to the Technical Specifications Date of Issuance:
November 6, 1991
ATTACHMENT TO LICENSE AMENDMENT NO. 51 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.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 2-2 2-2 2-4 2-4 3/4 2-9 3/4 2-9 B 2-1 B 2-1 B 2-2 B22 B 3/4 2-4 B 3/4 2-4 8 3/4 2-6 B 3/4 2-6 B 3/4 4-1 B 3/4 4-1 r
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS P
2.1 SAFETY LIMITS REACTOR CORE i
2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T3yg) shall not exceed the limits shown in Figure 2.1-1 for four i sop operation.
APPLICABIL1TY:
MODES 1 and 2.
ACTION:
3 Whenever the point defined by the combination of the highest operating loop average temperature and THERRAL PCWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY withirrl hour, and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
WhenevertheReactorCoolantSystempressurehasexceeded2735psiglimit be in HOT STANDBY with the Reactor Coolant System pressure withir, its l
wtWiin.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
MODES 3, 4, and 5:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.
i WOLF CREEK - UNIT 1 2-1
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0.0 0.1 0.2 0.3 0.4 0.5
- 0. f, 0.7 0.8 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.) 1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION WOLF CREEK - UNIT 1 2-2 Amendment No, al
TABLE 2.2-1 g8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m
9 p
SENSOR z
TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)
Z_
(S)
TRIP SETPOINT ALLOWABLE VALUE 1.
Marrial Reactor Trip N.A.
N.A.
N.A.
N.A.
N.A.
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2.
Power Range, Neutron Flux a.
High Setpoint 7.5 4.56 0
1109% of RTP*
<112.3% of RTP*
b.
Low Setpoint
'8.3 4.56 0
<25% of RTP*
<28.3% of RTP*
3.
Power Range, Neutron Flux, 2.4 0.5 0
<3% of RTP* with
<6.3% of RTP* with High Positive Rate i tinse csnstant i ti:ne constant 12 seconds 12 seconds 4.
Power Range, Neutron Flux, 2.4 0.5 0
<4% of RTP* with
<6 3% of RTP* with High Negative Rate i time constant i time constant ry 12 seconds
>2 seconds s
5.
Intermediate Range, 17.0 8.41 0
125% c. RTP*
135.3% of RTP*
Neutron Flux 5
5 6.
Source Range, Neut on Flux 17.0 10.01 0
$10 ces 11.6 x 10 cps 7.
Overtemperature aT
- 7. 2 3.40 2.49 See Note 1 See Note 2 f
g 8.
Overpower AT 5.5 1.43 0.15 See Note 3 See Note 4 5.
Pressurizer Pressure-Low 3.7 0.71 2.49 11915 psig 11906 psig j
5 10.
Pressurizer Pressure-High
- 7. 5 0.71 2.49
<2385 psig 12400 psig 11.
Pressurizer Water Level-High 8.0 2.18 1.95
<92% of instrument
<93.9% of instrument y
span span O *RTP = RATED THERML POWER
- Loop design flow = 93,750 gpm
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i MEASUREMENT UNCERTAINTIES OF 2.5% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF FJ ARE INCLUDED IN THIS FIGURE 48 46 ACCEPTABL2 UNA00tPTABLE c;PEMDON OPtnATM Aso m.
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0.90 0.95 1.00 1.05 1.10 R = M/1.49[1.0 + 0.3(1.0 P)]
FIGURE 3.2-3 RCS TO'iAL FLOW RATE VERSUS R FOUR LOOPS IN OPERATION
-i WOLF CREEK - UNIT 1 3/4 2-9 Amendment No.23, 51
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POWER DISTRIBUTION LIMITS 7
LIMITING CONDITION FOR OPERATION 2=
ACTION'(ConHhead);
- . s c.
Identify and correct theicause of the out-of-limit condition prior-to increasing THERMAL POWER above the reduced THERMAL POWER limit required by-ACTION a.2.=and/or b., above; subsequent POWER OPERATION may procet> provided that the combination of R and indicated RCS i.
total flo ate are _ demonstrated, through incore flux mapping and RCS total-flow rate comparison, to be within the region of_ acceptable s
m operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:
1, A nominal 50% of. RATED THERMAL POWER, c
2.
-;A: nominal 75% of RATED THERMAL POWER, and F
i 3.
Within 241 hours0.00279 days <br />0.0669 hours <br />3.984788e-4 weeks <br />9.17005e-5 months <br /> of attaining greater than or equal to 95% of ~
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RATED THERMAL-POWER.
e SURVEILU.14CE REQUIREMENTS
- 4.2.3.1 The provisions of Specification 4.0.4-are not applicable.
4.2.3.2. The combination of indicat6d RCS-total-flow rate and R shall be determined to_be within the region of acceptable operation of Figure 3.2-3:
a.-
Prior to operation above 75% of RATED' THERMAL POWER after each fuel loading, and b.
At least once per 31 Effective Full Power. Days.
14.2.3.3! The indicated RCS total flow rate shall be verified to-be within the region of_ acceptable operation of_ tigure 3.2-3 at least<once per n i.ours when the most recently obtained value of R obtained per Specification 4.2.3.2, is assumed to exist.
.4.2.3.4.TheRCStotal-flowrateindicatorsshallbesubjectedtoaCHANNEL CALIBRATION at-least once per 18 months.
4.2.3.5 The RCS= total flow rate shall-be determined by precision heat balance weasurement at least once per 18 months.
Within 7 days prior to performing the. precision heat balance, the instrumentation used for determination-of steam
- pressure, feedwater pressure, feedwater temperature, and _feedwater venturi-q AP in= the calorimetric calculations shall be calibrated.
a 4.2.3.6 The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months.
i WOLF CREEK.- UNIT 1-3/4 2-10 1
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2.1 SAFETY LIMI1S BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of_ the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat t*ansfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been relat)d to DNB through DNBR correlations.
DNBR correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.
The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.
e The DNB design basis is as follows:
there must be at least a 95 percent probability that the minimum DNBR of the N iting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application).
The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will rot occur when the minimum DNBR is at the DNBR limit (1,17 for the WRB-1 correlation).
For plant conditions which fall outside the range of applicability of the WRB-1 correlation, the W-3 correlation is used.
In addition, DNB margin is maintained by performing safety analyses to a higher valve than the correlation limit, called the safety analysis limit DNBR.
The margin between the safety analysis limit DNBR and the correlation limit DNBR is used to cover known DNBR penalties and provide margin for design flexibility.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the applicable safety analysis limit DNBR, or the average l
enthalpy at the vessel exit is equal to the anthalpy of saturated liquid.
These curves are based on an enthalpy hot channel factor, F f 1.55 H,
and a reference cosine with a peak of 1.55 for axial power shape.
An allowance isincludedforanincreaseinFfH at reduced power based on the expression:
F H = 1.55 [1+ 0.3 (1-P)]
Where P is the fraction of RATED THERMAL POWER.
WOLF CREEK - UNIT 1 B 2-1 Amendment No. 23, 51
= _ -
SAFETY LIMITS
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B SES-2.1.1 RFACTOR CORE (Continued) l These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the ft (AI) function of the Overtempert N n trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor vessel, pressurizer, and the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with tha design criteria and associated Code requirements.
The entire RCS is hydrotested at greater than or equal to 125% (3110 psig) of design pressure, to demonstrate integrity prior-to initial operation.
WOLF CREEK - UNIT 1 B 2-2 Amendment No. 51
l POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACIDR (Contirued) c.
The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
N F
will be maintained within its limits provided Conditions a. through g
AsnotedonFigure3.2-3,RCSflowrateandFfg
- d. above are maintained.
may be " traded off" against one another (i.e., a low measured RCS flow N
rate is acceptable if the measured F is also low) to ensure that the calcu-g lated DNBR will not be below the safety analysis DNBR value.
The relaxation ofFhasafunctionofTHERMALPOWERallowschangesintheradialpowershape for all permissible rod insertion limits.
R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts N
for F less than or equal to 1.49.
This value is used in the various accident analyses where F influences parameters other than DNBR, e.g., peak clad tem-g perature, and thus is the maximum "as measured" value allowed.
Fuel rod bowing reduces the value of DNB ratio.
Credit is available to offset this reduction in the generic margin.
The generic margins, totaling l
11.4% DNBR, completely of fset any rod bow penalties.
This is the margin between the correlation DNBR limit (1.17) and the safety analysis limit DNBR (1,32).
The applicable values of rod bow penalties are referenced in the FSAR.
When an F measurement is taken, an allowance for both experimental error q
and manufacturing tolerance must be made.
An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
WOLF CREEK - UNIT 1 B 3/4 2-4 Amendment No. 51
. - ~ _ _ _ _
4 POWER DISTRIBUTION LIMITS BASES
=-
flEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR FNTHALPY RlSE HOT CHANNEL FACTOR (Continued)
The Radial Peaking factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (z) remains within its limit.
The F
limit for RATED THERMAL POWER (F P)9 as provided in the Radial Peaking xy Factor Limit Report per Specification 6.9.1.9 was determined from expected power control manuevers over the full range of burnup conditions in W core.
WhenRCSflowrateandFfH are measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-3.
Measurement errorsof2.5%forRCStotalflowrateand4%forF[H have been allowed for in chtermination of the design DNBR value.
i e
The measurement error for RCS total flow rate is based upon perfaming a precision heat balance and using the result to calibrate the RCS flow rate
=
indicators.
Potential fouling of the feedwater venture which might not be l
detected could bias the result from the precision heat balance in a non-j conservative manner.
Therefore, an inspection is performed of the feedwater j
venture each refueling outage.
i The 12-hour periodic surveiliance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3.
This surveillance also provides adequate monitoring to detect any core crud buildup, f
3/4.2.A QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis, Radial power distribution measurements are made during STARTUP testing and i
periodically during power operation.
t The limit of 1.02, at which corrective ACTION is required, provides DNB and linear heat generation rate protection with x y plane power tilts.
A limit.
of 1.02 was selected to provide an allowance for the uncertainty associated with l
the indicated power tilt.
1 The 2-hour time allowance for operation with a tilt condition greater-than 1.02 but less than 1.09 is provided to allow identitication and correc-tion of a dropped or misaligned control rod.
In the event such ACTION does l
not' correct the tilt, the margin for uncertainty on F is reinstated by reducing q
j the maximum allowed power by 3% for each percent of tilt in excess of 1.
I i
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WOLF CREEK - UNIT 1 B 3/4 2-5 Amendment No. 23 4
=.
POWER DISTRIBUTION LIMITS BASES QUADRANTPOWERTILTRATIO(Continued)
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore
-detector is inoperable, the moveable incore detectors are used to confirm that the normalized-symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
Tha incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
The two sets of four symmetric thimbles is a unique set of eight detector locations.
These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
3/4.2.5 DNR PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of opetation assumed in the transient and accident analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a DNBR-above the safety analysis limit DNBR (1.32) throughout each analyzed transient.
The indicated T valve of 592.5 F and the indicated avg
_ pressurizer pressure value of 2220 psig correspond to analytical limits of 595 F and 2205 psig-respectively, with allowance for measurement uncertainty.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
WOLF CREEK - UNIT 1 B 3/4 2-6 Amendment No. 51
=
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4l1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the safety analysis limit DNBR (1.32) during all normal operations and anticipated transients.
In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident; however, single failure considerations require that three loops be OPERABLE.
A single reactor coolant loop provides sufficient heat removal if a bank withdrawal accident can be prevented; i.e., by opening the Reactor Trip System breakers.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE, In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a reactor coolant pump.in MODES 4 and 5 are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steare generator is less than 50 F above each of the RCS cold leg temperatures.
3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam.
The relief capacity of a single' safety valve is adequate to relieve any overpressure condition which could_ occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.
l WOLF CREEK - UNIT 1 B 3/4 4-1 Amendment No. 51
.. _~
3/4.4 REACTOR COOLANT SYSTEM PASES SAFETY VALVES (Continued)
During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip and also assuming no operation of the power-operated relief valves or steam dump vasves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of S.rtion XI of the ASME Boiler and Pressure Code.
3/4.4.3 -PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation.
The maximum water _ volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.
The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant-System pressure and establish natural circulation.
3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including tha design step load-decrease with steam dump.
Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
Each PORV has a remotely operated block valve to provide-a positive shutoff capability should a' relief valve become inoperable.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-l tained. The program for inservice = inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveil-1 lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so'that corrective measures can be taken.
l-WOLF CRE:K - UNIT 1 B 3/4 4-2 l
-,