ML20236M486
| ML20236M486 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/04/1987 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236M488 | List: |
| References | |
| TAC-65318, NUDOCS 8708110085 | |
| Download: ML20236M486 (13) | |
Text
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[faarogjo, UNITED STATES j.
i NUCLE AR REGULATORY COMMISSION l
y WASHINGTON, D. C. 20555 j
.....f i
KANSAS GAS & ELECTRIC COMPANY KANSAS CITY POWER AND LIGHT COMPANY KANSAS ELECTRIC POWER COOPERATIVE, INC.
WOLF CREEK GENERATING STATION l
DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 9 l
License No. NPF-42 l
l 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 Kansas Gas and Electric Company acting for itself and Kansas City Power and Light Company and Kansas Electric Power Cooperative, Inc., (licensees) dated May 7,1987, complies with the standards and requirements of j
the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; l
l 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.
8708110085 870804 PDR ADDCK 05000482 P
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2.
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 License No. NPF-42 is hereby amended.to read as follows:
2.
Technical Specification The Technical Specifications contained in Appendix A, as revised through Amendment No. 9, and the Environmental Protection Plan
- 3 contained in Appendix B, both of which are attached hereto, are
]
hereby incorporated in the license.
KG&E shall 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.
I FOR THE NUCLEAR REGULATORY COMMISSION dc
&ds-o
"/p Jose A. Calvo, Director Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects l
Office of Nuclear Reactor Regulation
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Attachment:
Changes to the Technical Specifications Date of Issuance: August 4, 1987 l
j
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l ATTACHMENT TO LICENSE AMENDMENT NO. 9 FACILITY 0;; RATING 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 page is also provided to maintain document completeness, REMOVE PAGES INSERT PAGES 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-33 3/4 3-33 i
3/4 3-34*
3.4 3-34*
B 3/4 3-1 B 3/4 3-1 B 3/4 3-2**
B 3/4 3-2**
B 3/4 3-3**
B 3/4 3-3**
l B 3/4 3-4**
B 3/4 3-4**
l B 3/4 3-5**
B 3/4 3-5**
B 3/4 3-6**
B 3/4 3-6**
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- 0verleaf Page
- Pages reissued due to repagination 1
l e
TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS j
l 1.
Manual Initiation a.
Safety Injection (ECCS)
N.A.
b.
Containment Spray N.A.
c.
Phase "A" Isolation N.A.
d.
Phase "B" Isolation N.A.
e.
Containment Purge Isolation N.A.
f.
Steam Line Isolation N.A.
j I
g.
Feedwater Isolation N.A.
h.
Auxiliary Feedwater N.A.
i.
Essential Service Water N.A.
j Containment Cooling N.A.
j k.
Control Room Isolation N.A.
l.
Reactor Trip N.A.
)
m.
Emergency Diesel Generators N.A.
i n.
Component Cooling Water N.A.
o.
Turbine Trip N.A.
i l
2.
Containment Pressure-High-1 a.
Safety Injection (ECCS) 5 29(7)/27(4) l t
1)
Reactor Trip 52 j
2)
Feedwater Isolation
<7
)
3)
Phase "A" Isolation 5 1.5(5) 4)
< 60 5)
Essential Service Water
[60(1)
)
6)
Containment Cooling 5 60(1)
]
l 7)
Component Cooling Water N.A.
8)
Emergency Diesel Generators 514(6) 9)
Turbine Trip N.A.
l i
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WOLF CREEK - UNIT 1 3/4 3-29 Amendment No. 9 i
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TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME'IN SECONDS 3.
Pressurizer Pressure-Low a.
Safety Injection (ECCS)
'$ 290)/27(4)
I 1)
Reactor Trip 52 l
2)
Feedwater Isolation
<7 3)
Phase "A" Isolation f2(5) l 4)
< 60 5)
Essential Service Water 60(1) 6)
Containment Cooling 5 60(1) l 7)
Component Cooling Water N.A.
l 8)
Emergency Diesel Generators 5 14(6) 9)
Turbine Trip N.A.
4.
Steam Line Pressure-Low I
a.
SafetyInjection(ECCS)
$39(3)/27(4) l 1)
Reactor Trip 52 j
2)
Feedwater Isolation
<7 I
[2(5) 3)
Phase "A" Isolation 4)
< 60 5)
Essential Service Water 60(1) 6)
Containment Cooling-560(1) l 7)
Component Cooling Water N.A.
8)
Emergency Diesel Generators 514(6) 9)
Turbine Trip N.A.
b.
Steam Line Isolation 5 2(5)
WOLF CREEK - UNIT 1 3/4 3-30 Amendment No. 9
4 4
d L>
s TABLE 3.3-5 (Continued) l TABLE NOTATIONS 4
(1) Diesel generator starting and sequence loading delays included.
)
)
t' (2) Diesel generator starting delay not included.
Offsite power available.
i (3) Diesel generator starting and sequence loading delay included.
RHR pumps not included.
Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves'open, then VCT valves close) is included.
(4) Diesel generator starting and sequence loading delays not included.
Offsite power available.
RHR pumps'not included.
Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is included, g(5)
Does not include valve closure time.
(6)
Ir.cludes time for diesel to reach full speed.
(7) Diesel generator starting and sequence loading delays included.
Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is not included.
Response time assumes only opening-of RWST valves.
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3/4.3 INSTRUMENTATION BASES j
l 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION Sis 7ETIETRUMENTATION The OPERABIllT) df the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensure that:
(1) the I
associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each cAaiinel or combination thereof reaches its Setpoint, (2) the specified coincidence logic !s maintained, (3) sufficient redundancy is main-I tained to permit a channe? m be out-of-service for testing or maintenance, and l
(4) sufficient system functional. capability is available from diverse parameters.
W e OPERABILITY of these systems is required to provide the overall reli-ability, redundancy, and diversity assumed available in the facility design for the protection and'nitigation of accident and t*ansient conditions.
The inte-grated operation of each of these systems is consistent with the assumptions used in the safet9 analyses.
The Surveillance Requirements specified for these l
systems ensure that the overall system functional capability is maintained com-i parable to the original design standards.
The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability.
The Engineered Safety Features Actuation System Instrumentation Trip Set-points specified in Table 3.3-4 are the nominal values at which the bistables are set for each fun tional unit.
ASetpointisconsideredtobeadjustedcon-sistent with the nominal vMue when the 'as measured" Setpoint is within the 4
band allowed for calibration accuracy.
i i
I i
ESF response times specified in Table 3.3-5 which include sequential opera-tion of the RWST and VCT valves (Notes 3 and 4) are based on values assumed in 1
the non-LOCA safety analysis.
These analyses take credit for injection of l
borated water from the RWST.
Injection of borated waterfi+ assumed not to occur until the VCT charging pump suction valves are clohed foll ving opening of the RWST charging pump suction valves. When the sequential operation of the RWST and VCT valves is not included in the response times (Not?. 7), the values specified are based on the LOCA analyses.
the LOCA analyses tWe credit for injection flow regardless of the source.
Verification of the response times l
1 specified in Table 3.3-5 will assure that the assumptions used for the LOCA and l
l non-LOCA analyses with respect to operation ~of the VCT and RWST valves are valid.
l i
1,
To accommodate the instrument drift assumed to occur t d ween operational l
tests and the accuracy to which Setpoints can be measured and calibrated, i
Allowable Values for the Setpoints have been specified in Table 3.3-4.
Opera-l tion 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 accommod3te this error.
An optiotal provision has been included for determining the Operability of a channel when its Trip Setpoint is found to exceed the deviation from the specified calibration point,fpr rack and sensor components 4 conjunction with a stat 9 tical coabinatisn of the other uncertainties in calibrating the instrumentagon.
In Equat W 9 3 3-1, Z + R + 5
. TA, the interactive effects of the errors in'the rack and the sensor, and the l
WOLF CREEK - UNIT 1 B 3/4 3-1 k
Amendment No. 9
l 1
4 3'4.3 INSTRUMENTATION BASES 1
3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
"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 the difference, I
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.
S or Sensor Error is either tha "as measured" deviation of the sensor from its calibration point or the valr: 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 l
will happen, an infrequent excessive drift is expected.
Rack or sensor drif t,
)
in excess of the allowance that is more than occasional, may be indicative of I
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 associated with each channel is completed within the time limit assumed in the l
safety analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.
i Sensor reponse time verification may be demonstrated by either:
(1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors l
with certified response times.
1 The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being ex-ceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients.
Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the. conditions.
As an example, the following actions say be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break break or loss-of-coolant accident:
(1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) Feedwater System isolates, (4) the emergency diesel generators start, (5) containment spray pumps start.and automatic valves position, (6) containment isolates, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater WOLF CREEK - UNIT 1 B 3/4 3-2
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h L_____-__-____
e I
INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES XCTUATION SYSTEM INSTRUMENTATION (Continued) pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.
Engineered Safety Features Actuation System Interlocks The Engineered Safety Features Actuation System interlocks perform the following functions:
P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T below Setpoint, prevents the opening of the main avg feedwater valves which were closed by a Safety Injection block so that components can be reset or tripped.
Reactor not tripped prevents manual block of Safety Injection.
P-11 On increasing pressure P-11 automatically reinstates ' safety injec-tion actuation on low pressurizer pressure and low steamline pressure and automatically blocks steamline isolation on negative steamline pressure rate.
On decreasing pressure; P-11 allows the manual block or Safety Injection on low pressurizer pressure and l
low steamline pressure and allows steamline isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that:
(1) the associated ACTION will be initiated when the raidation level monitored by each channel or combinatin thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for-testing or maintenance.
The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions.
Once the required logic com-bination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and acutation of Emergency Exhaust or Control Room Emergency Ventilation Systems.
3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment enssures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the WOLF CREEK - UNIT 1 B 3/4 3-3
l l
4 INSTRUMENTATION l
BASES l
i 3/4.3.3.2 MOVABLE INCORE DETECTORS (Continued) core.
The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of iis voltage curve.
l N
I For the purpose of measuring F (Z) or vH a full incore flux map is used.
A I
Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore l
flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT l
POWER TILT RATIO when one Power Range Neutron Flux channel is inoperable.
l 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event i
and evaluate the response of those features important to safety.
This capabil-ity is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100.
The instrumentation is consistent l
with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earth-quakes," April 1974.
3/4.3.3.4 METEOROLOGICALINSTRUMENTAlIOJ!
The OPERABILITY of the meteorological instrumentation 2nsures that suf-ficient meteorological data is available for estimating potential radiatior doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.
This capability is required to evaluate the need for initiating protective measures to protect the healthe and safety of the ublic and is cnnsistent with the recommendations of Regulatory Guide 1.23, p'Onsite Meteorological Programt," February 1972.
3/4.3.3.5 REMOTE SHUTOOWN INSTRUMENTATION The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the facility from locations outside of the control room and that a fire will not preclude achieving safe shutdown.
The Remote Shutdown System transfer switches, power circuits, and control circuits are independent of areas where a fire could damage systems normally sued to shut down the reactor.
This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 3 and 19 and Appendix R of 10 CFR Part 50.
3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.
This capability is cen-sistent with the recommendations of Regulatory Guide 1.97, Revision 2, WOLF CREEK - UNIT 1 B 3/4 3-4
INSTRUMENTATION BASES 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION (Continued)
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant i
Conditions During and Following an Accident," December 1980 and NUREG-0737,
" Clarification of TMI Action Plan Requirements," November 1980.
3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection System ensures that sufficient capability is available to promptly detect and intiate protective action in the event of an accidental chlorine release.
This capability is required to protect control and room personnel and is consistent with the recommendations i
of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," Revision 1, January 1977.
3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that both ade-quate warning capability is available for the prompt detection of fires and that Fire Suppression Systems, that are actuated by fire detectors, will discharge extinguishing agents in a timely manner.
Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than i
detectors tnat are installed solely for early fire warning and notification.
Consequently, the minimum number of operable fire detectors must be greater.
The loss of detection capability for Fire Suppression Systems, actuated by fire detectors represents a significant degradation of fire protectkn for any area.
As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning.
The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.
3/4.3.3.9 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System areconsistentwiththerecommendationsofRegulatoryGuide1.133,guirements components.
The allowable out-of-service times and Surveillance Re
' Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.
3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The Alarm /
WOLF CREEK - UNIT 1 B 3/4 3-5
INSTRUMENTATION l
BASES 3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION (Continued)
Trip Setpoints for these instruments shall be calculated and adjusted in l
accordance with the methodology and parameters in the ODCM to ensure that the l
alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
3/4.3.3.11 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous i
effluents during actual or potential releases of gaseous effluents.
The f
Alarm / Trip Setpoints for these instruments shall be adjusted to values calculated l
in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The instrumentation also includes provisions fe" monitoring (and controlling) the I
concentrations of potentially explosive gas mixtures in the WASTE GAS HOLOUP l
SYSTEM.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
The sensitivity of any noble gas activity monitor used to show compli-ance with the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10 G pCi/cc are measurable.
3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed pro-tection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.
Although the orientation of the turbine is such that the number of potentially damaging missiles which could impact and damage safety-reinted components, equipment, or structures is minimal, protection from excessive turbine overspeed is required.
WOLF CREEK - UNIT 1 B 3/4 3-6