ML17285B239
| ML17285B239 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/20/1990 |
| From: | Chan T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17285B240 | List: |
| References | |
| NUDOCS 9005030237 | |
| Download: ML17285B239 (12) | |
Text
gpS RECT
+s
~
0 I ~
I Q
I g
g I
a~'v go
++*<<+
UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 WASHINGTON PUBLIC POWER SUPPLY. SYSTEM DOCKET NO. 50-397 NUCLEAR PROJECT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 8>
License No. NPF-21 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
B.
C.
D.
E.
The application for amendment filed by the Washington Public Power Supply System (the licensee),
dated November 29, 1989 and supplemented by letter dated March 21, 1990 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; 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 set forth in 10 CFR Chapter I; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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.
'9005030237
'900420 PDR ADOCK 05000397 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-21 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
81, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective as of the date of issuance.
FOR TH U
EAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 20, 1990 Te ence L. Chan, Acting Director Project Directorate V
Division of Reactor Projects - III, IV, V and Special Projects
ENCLOSURE TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO.
NPF-21 DOCKET NO. 50-397 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Also to be replaced are the following overleaf pages.
AMENDMENT PAGE 3/4 3-56 3/4 3-57 3'4 3-4
- B , ~ 3-5
- B 3/4 3-6 OVERLEAF PAGE 3/4 3-55 3/4 3-58 B 3/4 3-3
- The text on these pages is shifted but no change is made to the content of the text.
~
V I
INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occur-rence of a failure to scram during an anticipated transient.
The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, and NE00-24222, dated December 1979.
The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of the reactor protection system and is an essential safety supplement to the reactor trip.
The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle.
The physical phenomenon involved is that the void ~eactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity.
Each EOC-RPT system trips both recircula" tion pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events.
The two events for which the EOC-RPT protective feature will function are closure of the turbine throttle valves and fast closure of the turbine governor valves.
A fast closure sensor from each of two turbine governor valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine governor valves provides input to the second EOC-RPT system.
Similarly, a position switch for each of two turbine throttle valves provides input to one 'EOC-RPT system; a position switch from each of the other two throttle valves provides input to the other EOC-RPT system.
For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine governor valves and a 2-out-of-2 logic for the turbine throttle valves.
The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.
Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled.
The manual bypasses and the automatic Operating 8ypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.
The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,
- 190ms, less the time allotted for sensor response, i.e.,
- 10ms, and less the time allotted for breaker arc suppression determined by test, as correlated to.manufacturer's test results, i.e., 83ms, and plant preoperational test results.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
WASHINGTON NUCLEAR - UNIT 2 8 3/4 3-3
INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of Specifications 3/4.1.4, Control Rod Program Controls, 3/4.2, Power Distribution Limits and 3/4.3. 1 Reactor Protection System Instrumentation.
The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The test exception to the weekly Channel Functional Test of the SRM/IRM Detector Not Full In instrumentation noted in Table 4.3.6-1, Control Rod Block Instrumentation Requirements, is intended to avoid cable damage and radiation exposure during operational condition 5 periods when outage work is being done in the under core region.
Upon completion of all the work in this area, when access for maintenance or construction efforts is no longer required, the test will be completed per the prescribed frequency within seven days.
3/4.3. 7 MONITORING INSTRUMENTATION 3/4. 3. 7. 1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels; (2) the alarm is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.
This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.
The criticality monitor alarm setpoints were calculated using the criteria from 10 CFR 70.24.a.1 that requires detecting a dose rate of 20 Rads per minute of combined neutron and gamma radiation at 2 meters.
The alarm setpoint was determined by calculational methods using the gamma to gamma plus neutron ratios from ANSI/ANS 8.3-1979, Criticality Accident Alarm System, Appendix B and assuming a critical mass was formed from a seismic event, with a volume of 6' 6' 6't a distance of 27.7 feet from the two detectors.
The calculated dose rate using the methodology is 5.05 R/hr.
The allowable value for the alarm setpoint was, therefore, established at 5R/hr.
WASHINGTON NUCLEAR - UNIT 2 B 3/4 3-4 Amendment No. 81
INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued)
- 3. 4. 3. 7. 2 SEISMIC MONITORING INSTRUMEHTATIOH The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.
This capability is required to permit comparison of the measured response to that used in the design basis for the unit.
This instrumentation is consistent with the recommendations of Regulatory Guide 1. 12, "Instrumentation for Earthquakes,"
April 1974.
3/4. 3. 7. 3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data are available for estimating potential radiation 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 health and safety of the public.
This instrumentation is consistept with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs,"
- February, 1972.
3/4. 3. 7. 4 REMOTE SHUTDOWN MONITORING INSTRUMENTATIOH The OPERABILITY of the remote shutdown monitoring instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room.
This capability is required in the 'event control room habitability is lost and is consistent with General Design Criterion. 19 of Appendix A to 10 CFR Part 50.
3/4. 3. 7. 5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.
This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, "Clarification cf TMI Action Plan Requirements,"
November 1980.
3/4. 3. 7. 6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown.
At these power levels, reactivity additions shall not be made without this flux level information available to the operator.
When the inter-mediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.
3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.
WASHINGTON NUCLEAR - UNIT 2 B 3/4 3-5 Amendment No.
8>
INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.9 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components.
The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1. 133, "Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors,"
May 1981.
3/4. 3. 7. 10 RADIOACTIVE LI UID 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/trip setpoints for these instruments shall be calculated and adjusted 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 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 purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.
WASHINGTON NUCLEAR " UNIT 2 B 3/4 3-6 Amendment,No.
81
TRIP FUNCTION TABLE 3.3.6-2 CONTROL ROO BLOCK INSTRUHENTATION SETPOINTS TRIP SETPOINT ALLOMABLEVALUE 1.
2.
3.
5.
6.
ROD BLOCK HONITOR b.
Inoperative c.
Downscale APN a.
Flow Biased Neutron Flux Upscale b.
Inoperative c.
Downscale d.
Neutron Flux " Upscale, Startup SOURCE RANGE NNITORS a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale INTERMEDIATE RANGE NNITORS a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale SCRN DISCHARGE VOLTE a.
Mater Level-High b.
Scram Trip Bypass REACTOR COOLANT SYSTN RECIRCULATION FLSt
< 0.66 M+ 40K R.A.
> 5% of RATED THERNL POMER
< 0.66 1+ le~
R.A.
> 5X of RATED THERNL POMER
< 12K of RATED THERNAL POMER N.A.
< 1 x 10 cps H.A.
> O.T cps N.A.
< 108/125 divisions of full scale H.A.
> 5/125.divisions of full scale
< 527 ft 2 in. elevation H.A.
< 0.66 N+ 43%
H.A.
> 3X of RATED THENAL POMER t
< Oe66 M+ l%
Q.A.
> 3L of RREED THERIRL ENER ~
< 14% of RATED THERNL POMER N.A.
< 1.6 x 10 cps
.I.A.
> 0.5 cps N.A.
< 110/125 divisions of full scale R.A.
> 3/125 divisions of full scale c 557 ftl in. elevetIon S.A.
a.
Upscale
< 108/125 divisions of full scale N.A.
< 10K flow deviation on is varied as a function of eainta)ned in accordance with b.
Inoperative c.
Coeparator e
verage ower ange Honitor rod block functi (M).
The trip setting of this function Nust be
< 111/125 divisions of full scale II.A.
< 11K flan deviation recirculation loop flow Specification 3.2.2.
TABLE 4.3.6"1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP FUNCTION ROD BLOCK MONITOR a.
Upscal e b.
Inoperative c.
Downscale CHANNEL CHECK N.A.
N.A.
N.A.
CHANNEL FUNCTIONAL TEST S/U(b)(c), M(c)
S/U(b)(c), M(c)
S/U(b)(c), M(c)
CHANNEL CALIBRATION Q
N.A.
OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE RE UIRED 2.
3.
4.
5.
6.
APRM a.
Flow Biased Neutron Flux Upscale b.
Inoperative c.
Downscale d.
Neutron Flux - Upscale, Startup SOURCE RANGE MONITORS a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale INTERMEDIATE RANGE MONITORS a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale SCRAM DISCHARGE VOLUME a.
Water Level-High b.
Scram Trip Bypass REACTOR COOLANT SYSTEM RECIRCULATION a.
Upscale b.
Inoperative c.
Comparator N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
FLOW N.A.
N.A.
N.A.
S/U(b),
M S/U(b),
M S/u(b)',
M S/U(b),
M S/u(b), W(¹)
S/U(b)$
W S/U(b),
W S/U(b),
W S/U(b),
W S/U(b),
W S/U(b),
W S/U(b),
W S/U(b),
M S/U(b)',
M S/U(b),
M Q
N.A.
Q N.A.
Q N.A.
Q N.A.
Q N.A.
Q R
N.A.
N.A.
Q 1
1, 2, 5
1 2,
5 2,
5 2,
5 2,
5 2,
5 2,
5 2,
5 2,
5 2,
5 5A'A 5'A*
TABLE 4. 3. 6-1 (Continued)
CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE RE UIREMENTS TABLE NOTATIONS (a)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b)
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(c)
Includes reactor manual control multiplexing system input.
With THERMAL POWER
> 30K of RATED THERMAL POWER.
With more than one control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.1Q.1 or 3.9.10.2.
¹ This CHANNEL FUNCTIONAL TEST may be satisfied while in MODE 5 provided the detector is administratively controlled in the full in position and is visually verified to be in once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless the CHANNEL FUNCTIONAL TEST has not been performed within the past 92 days.
WASHINGTON NUCLEAR - UNIT 2 3/4 3"57 Amendment No. 81
INSTRUMENTATION 3/4.3. 7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE with their alarm setpoints within the specified limits.
APPLICABILITY:
As shown in Table 3.3.7.1-1.
ACTIDN:
ao b.
C.
kith a radiation monitoring instrumentation channel alarm set-point exceeding the value shown in Table 3.3.7.1-1, adjust the set-point to ~ithin the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or declare the channel inoperable.
With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.7.1-1.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE UIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL
- CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the conditions and at the frequencies shown in Table 4.3.7.1-1.
WASHINGTON NUCLEAR - UNIT 2 3/4 3-58 Amendment No. 74