ML20095E704
| ML20095E704 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/12/1995 |
| From: | Stone J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20095E706 | List: |
| References | |
| NUDOCS 9512180097 | |
| Download: ML20095E704 (13) | |
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t UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20665-0001
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WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 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 August 22, 1995, 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 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, ed (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this license amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
9512180097 951212 DR ADOCK 05000482 PDR
. 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 Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 93, 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 to be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION James C. Stone, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 12, 1995
ATTACHMENT TO LICENSE AMENDMENT NO. 93 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change. The correspondir.g overleaf pages are kiso provided to maintain document completeness.
REMOVE INSERT V
V 3/4 3-1 3/4 3-1 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-13 3/4 3-13 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-31 3/4 3-32 3/4 3-33 B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)..............................
3/4 2-1 3/4.2.2 HEAT FLUX HDT CHAIGGEL FACTOR - F,(X,Y,Z).................
3/4 2-4 3/4.2.3 NUCLEAR ENTHALPY RISE HDT CHANNEL FACTOR
-F(X,Y).............................................
3/4 2-9 m
3/4.2.4 QUADRANT POWER TILT RATI0................................
3/4 2-11 3/4.2.5 DNB PARAMETERS...........................................
3/4 2-14 TABLE 3.2-1 DNB PARAMETERS........................................
3/4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................
3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................
3/4 3-2 TABLE 3.3-2 DELETED l
TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE i
REQUIREMENTS........................................
3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 1
INSTRUMENTATION........................................
3/4 3-13 i
TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................
3/4 3-14 I
TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS......................
3/4 3-22 TABLE 3.3-5 DELETED I
TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SI'RVE I LLANCE REQUIREMENTS...........
3/4 3-34 WOLF CREEK - UNIT 1 V
Amendment No. E,93
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PXaE INSTRUMENTATIDN (Continued)
~
~
-3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................
3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS................................
3/4 3-40 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS........................................
3/4 3-42 Movabl e Incore Detectors.................................
DELETED l
Seismic Instrumentation..................................
DELETED TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION....................
DELETED TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
DELETED i
Meteorol ogical Instrumentation...........................
DELETED TABLE 3.3-B METEOROLOGICAL MONITORING INSTRUMENTATION.............
DELETED l
TABLE 4.3-5 METEOROL6GICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
DELETED I
r Remote Shutdown Instrumentation..........................
3/4 3-43 I
TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............
3/4 3-44 i
TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-45 l
Accident Monitoring Instrumentation......................
3/4 3-46 i
TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..................
3/4 3-47 l
TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-48 i
Chl ori ne Detection Systems...............................
DELETED Loose-Part Detection System..............................
DELETED l
Radioactive Liquid Effluent Monitoring Instrumentation.'..
DELETED TABLE 3.3-12 RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION....................................
DELETED l
WOLF CREEK - UNIT I VI Amendment No. 15,42,55,75, 89
. - -. ~. _.. _.. _ -.. - -. -. - - - -. - -. _ - ~. -. - - -.
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3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
l l
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
l As shown in Table 3.3-1.
SURVEILLANCE REOUIREMINTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the perfonnance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Neutron detectors are exempt from response time testing. Each test shall l
include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1, WOLF CREEK - UNIT 1 3/4 3-1 Amendment No. 93
_ TABLE 3.3-1 i
g REACTOR TRIP SYSTEM INSTRUMENTATION t
r-MINIMUM l
Po TOTAL NO.
CHANNELS CHANNELS APPLICABLE l
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FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.
1 2
1, 2 1
E 2
1 2
3*, 4*, 5*
10 Z
2.
Power Range, Neutron Flux a.
High Setpoint 4
2 3
1, 2 2#
b.
Low Setpoint 4
2 3
1#N, 2 2#
3.
Power Range, Neutron Flux 4
2 3
1, 2 2#
High Positive Rate 4.
Power Range, Neutron Flux, 4
2 3
1, 2 2#
High Negative Rate 5.
Intermediate Range, Neutron Flux 2
1 2
1M#, 2 3
E 6.
Source Range, Neutron Flux a.
Startup 2
1 2
2M**
4 b.
Shutdown 2
1 2
3**, 4, 5 5
7.
Overtemperature AT Four Loop Operation 4
2 3
1, 2 6#
8.
Overpower AT Four Loop Operation 4
2 3
1, 2 6#
9.
Pressurizer Pressure-Low 4
2' 3
1 6#
10.
Pressurizer Pressure-High 4
2 3
1, 2 6#
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WOLF CREEK - UNIT I 3/4 3-7 Amendment No. 93 (Next page is 3/4 3-9) i
i INSTRMENTATION 3/4 3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LINITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and < nterlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.
l APPLICABILITY: As shown in Table 3.3-3.
ACTION:
a.
With an ESFAS Instrumentation or Interlock Trip Satpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
b.
With an ESFAS Instrumentation or Interlock Trip Satpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, either:
1.
Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or 2.
Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + R + S s TA Where:
Z - The value from Column Z of Table 3.3-4 for the affected
- channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S - Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.
c.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
4 WOLF CREEK - UNIT 1 3/4 3-13 Amendment No. 84, 93
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THIS PAGE INTENTIONALLY BLANK WOLF CREEK - UNIT 1 3/4 3-29 Amendment No. 9,93 (Next page is 2/4 3-34)
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l 3/4.3-INSTRUMENTATION l
BASES 3/4.3.1~and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 1
The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensure that:
(1) the associated ACTION and/or Reactor trip willibe initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the ovaall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and' transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
When determining compliance with action statement requirements, addition to the RCS of borated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered to be a positive reactivity change.
The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, and Supplement 1, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," supplements to that report, and the NRC's Safety Evaluation dated February 21,1985, WCAP-10271 Supplement 2 and WCAP-10271-P-A Supplement 2, Revision 1, " Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," the NRC's Safety Evaluation dated February 22, 1989, and the NRC's Supplemental Safety Evaluation dated April 30, 1990.
Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability J
of the Reactor Protection System and Engineered Safety Features instrumentation.
I WOLF CREEK - UNIT I B 3/4 3-1 Amendment No. 0,12,43,93 2 n t r 22, 1003
INSTRUMENTATION i
BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLMENTATION (Continued)
I To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setmints can be measured and calibrated, Allowable Values for the Setpoints save been specified in Table 3.3-4.
Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equation 3.3-1, Z + R + S s TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered.
Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is l
the difference, in percent span, between the Trip Setpoint and the value used in the analysis for the actuation. R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. 5 or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation WOLF CREEK - UNIT 1 B 3/4 3-2 Amendment No. 48,93
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