ML20212F531
| ML20212F531 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  (NPF-42-A-113) |
| Issue date: | 10/20/1997 |
| From: | Thomas K NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212F535 | List: |
| References | |
| NUDOCS 9711040322 | |
| Download: ML20212F531 (11) | |
Text
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t UNITED STATES u -
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NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 30046 0001 WOLF CREEK NUCLEAR OPERATING CORPORATION tl0LF CREEK GENERATING STATIOS DOCKET NO. 50 482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 113 License No. NPF-42 1.
The Nuclear Regulatory Comission (the Comission) has found that:
1 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 July 3.1997, as supplemented by letter dated August 20. 1997 4
complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application, as amended, the provisiens of the Act, and the rules and regulations of the Comission:
i C.
There is reasonable a_ssurance:
(I) that t.he activities authorized by this amendn+nt cN 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 issuare. 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 Comicion's regulations and all applicable requirements have been satisfied.
971w4C322 971020 PDR ADGCK 05000402 P
' 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.113. 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 before restart from the ninth refueling outage currently scheduled to start on October 4. 1997.
FOR THE NUCLEAR REGULATORY COMMISSION W@W Kristine M. Thomas. 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:
October 20, 1997
t ATTACHMENT TO LICENSE AMENDMENT NO. 113 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Replace the following ) ages of the Appendix A Technical Specifications with the attached pages. T1e revised pages are identified tj Amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness, REM 0'!E INSERT 3/4 3-1 3/4 3-1 3/4 3-13a 3/4 3-13a B 3/4 3-2 B 3/4 3-2 B 2/4 3-3 B 3/4 3-3 t
t 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.
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1, SURVEILLANCE RE0VIREMENT3
~4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall bc demonstrated OPERABLE by the performance 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 tri) function shall be verified to be within its limit at least once per 18 montis.
Neutron detectors are exempt from response time testing.
Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that all channels are verified 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 Na, of Channels" column of Table 3.3-1.
WOLF CREEK - UNIT 1 3/4 3-1 Amendment No. 93.113
_ _ _ _ = _ - _
...s g
TABLE 3.3-1 q
REACTOR TRIP SYSTEM INSTRUMENTATION
.n.
i5lR MINIMUM TOTAL NO.
CHANNELS CHANNELS APPlICA8tE EUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE
. MODES ACTIM 1.
1 2
1,2 1
2 1
2 3*,4*,5*
10 2.
Power Range, Neutron Flux r.
High Setpoint 4
2 3
1,2 2f-b.
Low Setpoint-4 2
3 Ifff,2 2f l
R
- 3. -
Power Range, Neutron Flux, 4
2 3
1,2 2f j
High Positive pate 4.
Power Range, Neutron Flux, 4
2 3
1,2 ti High Negative Rate 5.
Intermediate Range, Neutron Flux 2
1 2
Ifff,2 3
6.
Source Range, Neutron Foux a.
Startup 2
1 2
2ff 4
b.
Shutdown 2
1 2
3,4,5 5
7.
Overtemperature AT 4
2 3
1,2 6#
Four Loop Operation 8.
Overpower AT 4
2 3
1,2 6f g
Four Loop Operation E+
9.
Pressurizer Pressure-Low 4
2 3
1 6
- 10. Pressurizer Pressure-High 4
2 3
1,2 6f 3'
D 8
__.______~
f
l-INSTRUMENTA(10N 3I42.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVElllANCE REQUIREMENTS 4.3.2.1' Each ESFAS. instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3 2.
4.3.2.2 The ENGINEERED SAFETY FEATURES P. ESPOUSE TIME
- of each ESFAS function shall be verified to be within.the limit at least once per 18 months.
Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that all channels are verified at<least once per N times 18 montns where N is the total number of redundant channels in a specific ESFAS function as shown in toe " Total No. of Channels" Column of Table 3,3-3.
- The provisions of Specification 4.0.4 are not applicable for response time testing of the steam turbine-driven auxiliary feedwater pump for entry into MODE 3.
WOLF CREEK - UNIT 1 3/4 3-13a Amendment No. 84,113
TABLE 3.3-3 I
6 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION G
MINIMUM c
E TOTAL NO.
CHANNELS CHANNELS APPLICABLE E
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
[
- 1. Safety Injection, (Reactor Trip, z
Pnase A" Isolation, Feedwater U
Isolation, Component Cooling j
Water. Turbine Trip. Auxiliary w
Feedwater-Motor-Driven Pump.
Emergency Diesel Generator Operation, Containment Cooling, and Essential Service Water Operation)-
a.
Manual Initiation 2
1 2
1,2,3,4 18 b.
Automatic Actuation Logic 2
1 2
1,2,3,4 14 and Actuation Relays q
,s (SSPS)
{
O c.
Containment Pressure-3 2
2 1,2,3 28*
l High-1 d.
Pressurizer Pressure-4 2
3 1, 2, 3#
28*
Low e.
Steam Line Pressure-Low 3/ steam line 2/ steam line 2/ steam line 1, 2, 3 28*
any stea2 line
- 2. Containment Spray a.
Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18
{
operated I
a simul-g l
taneously I
g z
b.
Autem tic Actuation 2
1 2
1,2,3,4 14 P
Logic and Actuation Relays (SSPS) c.
Containment Pressure-High-3 4
2 3
1,2,3 16 e
l
\\
3/4.3 INSTRUMENTATION BASFS 3/4.3.1 and 3/4.3.2 REACTO.B.IB1P SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 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 Reat. tor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic is maintained, (3) sufficient redundancy it 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 overall 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 staviards. The periodic surveillance tests performed at the minimum frequencin 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 reactivit/ 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 accadance with WCAP-10271, and Supplement 1, " Evaluation of Surveillance Frequencies and Out of Service Times for the D.eactor Protection Instrumentation System," supplements to that report, and the NRC's Safety Evaluation dated February 21, 1985, WCAP-IO271 Sipplement 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 H
times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.
I___.___-__-_____-_
WOLF CREEK - UNIT.1 B 3/4 3-1 Amendment No. 0,12,43,93 k = be 22, 1993 l
JNSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN$TRUMENTATION (Continued)
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Set)oints can be measured and calibrated.
Allowable Values for the Setpoints lave 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 arMysis to accommodate this error. An optional provision has been included w 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 3nalysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Triv Setpoint and the value used in the analysis for the actuation.
R or Rack Error is the "as measured" deviation in percent s)an, for the affected channel from the specified Trip Setpoint. S 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 :nalysis 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 ca)able of operating within the allowances of these uncertainty magnitudes.
Racc drift in exct.s 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 verification of response time at the specified frequencias provides assurance that the reactor trip and the engineered safety features actuation associated with ach channel is completed within the time limit assumed in the safety analysis.
No credit is taken in the analysis for those channels with response times indicated as not applicable (i.e.. N.A.).
Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the on the remainder of the channel, ponse times with actual response time tests summation of allocated sensor res Allocations for sensor response times may be obtained from:
(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests). (2) inplace, onsite, or offsite (e.g., vendur) test measurements. cr (3) utilizing vendor engineering specifications. WCAP-13o32-P-A Revision 2. " Elimination of Pressure Sensor WOLF CREEK - UNIT 1 B 3/4 3-2 Amendment No. 4 --93, 113 S verber 22. 1993
l INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
Response Time Testing Requirements" provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP.
Response
time verification for other sensor types must be demonstrated by test.
The allocation for sensor response times must tm verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time.
In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The Engineered Safety Features Actuation System senses selected plant parameters 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, events, and transients.
Once the required logic combination is completed, the system sends actrl~'
signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition.
As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line 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 ge,1erators start. (5) containment spray pumps start and automatic valves position. (6T containment isolates. (7) steam line isolation. (8) Turbine trip. (9) auxiliary feedwater 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.
Enaineered 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 feedwater. valves wWchwereclosedby7 Safety Injection or High Steam Generator Water Level signal, allows 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 injection 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 of Safety Injection on low pressurizer pressure and low steamline pressure and allows steamlime isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI.
WOLF CREEK - UNIT 1 B 3/4 3-3 Amendment No. 13,113 November 22. 1993 1
9 INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRINENTATION -
3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that:
(1) +,he associated ACTION will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation unitors for plant operatit,ns senses radiation levels in selected plant systens and locations and determines
. If they are, the whether or not :) redetermined limits are being exceeded.
signals are con)ined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Control Room Emergency Ventilation Systems.
3/4.3.3.2 DELETED 3 /4.3.3.3 DELETED WOLF CREEK - UNIT 1 B 3/4 3-4 Amendment No. G,39
_