ML20214Q868
| ML20214Q868 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/09/1986 |
| From: | Rood H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214Q871 | List: |
| References | |
| TAC-54724, TAC-54725, TAC-54866, TAC-54867, TAC-54868, TAC-54869, TAC-57301, TAC-57302, NUDOCS 8609260389 | |
| Download: ML20214Q868 (30) | |
Text
.. _ _ _ _ _ _ _ _ _ _ _ _ _ _
l.
l g,.a heuq$8,,
UNITED STATES g
NUCLEAR REGULATORY COMMISSION L
j WASHINGTON, D. C. 20555
%,...+/
SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY l
THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-361 SAN ON0FRE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. NPF-10 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment to the license for San Onofre Nuclear Generating Station, Unit 2 (the facility) filed by the Southern California Edison Company on behalf of itself and San Diego Gas and Electric Company, The City of Riverside and the City of Anaheim, California (licensees) dated April 2, and April 27, 1984 and March 18, and July 1, 1985, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as 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 regulations of the Comission; C.
There is reasonable assurance:
(1) that the activities authorized j
by this amendment can be conducted without endangering the health j
and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; 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; 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.
8609260389 860909 PDR ADOCK 05000361 Y
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this amendment and Paragraph 2.C(2) of Facility Operating License No. NPF-10 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 54, are hereby incorporated in the license.
SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment.
In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during change over shall be minimized.
4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h
y G; b t
-~
Harry Rood, Senior Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-B
Attachment:
Changes to the Technical Specifications Date of Issuance: September 9, 1986
September 9, 1986 ATTACHMENT TO LICENSE AMENDMENT NO. 54 FACILITY OPERATING LICENSE NO. NPF-10 DOCKET N0. 50-361 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendrent number and contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.
Amendment Pages Overleaf Pages 3/4 2-4 3/4 2-3 3/4 3-17 3/4 3-18 3/4 3-30 3/4 3-29 3/4 3-32 3/4 3-31 B 2-5 B 2-6 B 3/4 2-2 B 3/4 2-1
l POWER DISTRIBUTION LIMITS
(
1 3/4.2.3 AZIMUTHAL POWER TILT - To LIMITING CONDITION FOR OPERATION 3.2.3 The AZIMUTHAL POWER TILT (T shall be less than or equal to the AZIMUTHAL POWER TILT Allowance use8)in the Core Protection APPLICA8ILITY: MODE 1 above 20% of RATED THERMAL POWER.*
ACTION:
a.
With the measured AZIMUTHAL POWER TILT determined to exceed the AZIMUTHAL POWER TILT Allowance used in the CPCs but less than or equal to 0.10, within two hours either correct the power tilt or adjust the AZIMUTHAL POWER TILT Allowance used in the CPCs to greater than or equal to the measured value.
b.
With the measured AZIMUTHAL POWER TILT determined to exceed 0.10:
1.
Due to misalignment of either a part length or full length CEA, within 30 minutes verify that the Core Operating Limit
\\
Supervisory System (COLSS), when COLSS is being used to monitor r
the core power distribution per Specifications 4.2.1 and 4.2.4, is detecting the CEA misalignment.
2.
Verify that the AZIMUTHAL POWER TILT is within its limit within i
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Linear Power Level - High trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the AZIMUTHAL POWER TILT is verified within its limit at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified acceptable at 95% or greater RATED THERMAL POWER.
A See Special Test Exception 3.10.2.
i 1
i SAN ONOFRE-UNIT 2 3/4 2-3 l
i
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS i
i 4.2.3 The AZIMUTHAL POWER TILT shall be determined to be within the limit above 20% of RATED THERMAL POWER by:
a.
Continuously monitoring the tilt with COLSS when the COLSS is OPERABLE.
b.
Verifying at least once per 31 days, that the COLSS Azimuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT greater than the AZIMUTHAL POWER TILT Allowance used in the CPCs.
c.
Using the incore detectors at least once per 31 EFPD's to indepen-l dently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.
d.
Calculating the tilt at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the COLSS is inoperable.
SAN ONOFRE-UNIT 2 3/4 2-4 AMENDMENT NO. 54
2 TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATICN
?
hk MINIMUM
-d TOTAL NO.
CHANNELS CHANNELS APPLICABLE h)
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 9.
CONTROL ROOM ISOLATION (CRIS) a.
Manual CRIS (Trip Buttons) 2 1
1 All 13*#
b.
Manual SIAS (Trip Buttons) 2 sets of 2/ unit 1 set of 2 2 sets of 2/ unit 1, 2, 3, 4 8
c.
Airborne Radiation u,);
i.
Particulate / Iodine 2
1 1
All 13*#
ii. Gaseous 2
1 1
All 13*#
d.
Automatic Actuation U
Logic 1/ train 1
1 All 13*#
10.
T0XIC GAS ISOLATION (TGIS) a.
Manual (Trip Buttons) 2 1
1 All 14*#, 15*#
b.
Chlorine - High 2
1 1
All 14*#, 15*#
c.
Ammonia - High 2
1 1
All 14*#, 15*#
d.
Butane / Propane - High 2
1 1
All 14*#, 15*#
e.
Automatic Actuation Logic 1/ train 1
1 All 14*#, 15*#
5
f i
ll TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SR fn MINIMUM E
TOTAL NO.
CHANNELS CHANNELS APPLICABLE
((
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 11.
FUEL HANDLING ISOLATION (FHIS) a.
Manual (Trip Buttons) 2 1
1 16"#
b.
Airborne Radiation i.
Gaseous 2
1 1
16*#
ii. Particulate / Iodine 2
1 1
16*#
c.
Automatic Actuation Logic 1/ train 1
1 16*#
h 12.
CONTAINMENT PURGE ISOLATION (CPIS) w a.
Manual (Trip Buttons) 2 1
1 6
17b*#
b.
Airborne Radiation (2RT7804-1 or 2RT7807-2) i.
Gaseous 2
1 1
1,2,3,4 17a 6
17b*#
ii.
Particulate 2
1 1
1,2,3,4 17a 6
17b*#
iii. Iodine 2
1 1
6 17b*#
c.
Containment Area Radiation (Gamma) 2 1
1 1,2,3,4 17 (2RT7856-1 or 2RT7857-2) 6 17b*#
d.
Automatic Actuation Logic 1/ train 1
1 1,2,3,4 17 h
6 17b V 5
6-8 O
N
Table 3.3-5 (Continued)
(
INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 5.
Steam Generator Pressure - Low MSIS (1) Main Steam Isolation (HV8204, HV8205) 6.9 (2) Main Feedwater Isolation (HV4048, HV4052) 10.9 (3) Steam, Blowdown and Sample Isolation 20.9 l
(HV8419,HV8421)
(HV4053,HV4054,HV4057,HV4058)
(4) Auxiliary Feedwater Isolation (NOTE 7) 40.9 (HV4705, HV4713, HV4730, HV4731)
(HV4706, HV4712, HV4714, HV4715) 6.
Refueling Water Storage Tank - Low RAS (1) Containment Sump Valves Open 50.7*
7.
4.16 kv Emergency Bus Undervoltage LOV (loss of voltage and degraded voltage)
Figure 3.3-1 8.
Steam Generator Level - Low (and No
(
Pressure-Low Trip)
EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**
l (2) Auxiliary Feedwater (Steaan/DC train) 42.7 (NOTE 6) 9.
Steam Generator Level - Low (and AP - High),
EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**
l (2) Auxiliary Feedwater (Steam /DC train) 42.7 (NOTE 6) 10.
Control Room Ventilation Airborne Radiation CRIS (1) Control Room Ventilation - Emergency Mode Not Applicable 11.
Control Room Toxic Gas (Chlorine)
TGIS (1) Control Room Ventilation - Isolation Mode 16 (NOTE 5) 12.
Control Room Toxic Gas (Ammonia) l TGIS l
Control Room Ventilation - Isolation Mode 36 (NOTE 5)
SAN ONOFRE - UNIT 2 3/4 3-29 AMENDMENT NO. 46
Table 3.3-5 (Continued)
INITIATING SIGNAL ANO FUNCTION RESPONSE TIME (SEC) 13.
Control Room Toxic Gas (Butane / Propane)
TGIS Control Room Ventilation -
Isolation Mode 36 (NOTE 5) 14.
Fuel Handling Building Airborne Radiation FHIS Fuel Handling Building Post-Accident Cleanup Filter System Not Applicable 15.
Containment Airborne Radiation CPIS Containment Purge Isolation 2 (NOTE 2) 16.
Containment Area Dadiation CPIS Containment Purge Isolation 2 (NOTE 2)
NOTES:
1.
Response times include movement of valves and attainment of pump or blower discharge pressure as applicable.
2.
Response time includes emergency diesel generator starting delay (applicable to A.C. motor-operated valves other than containment purge valves), instrumentation and logic response only.
Refer to Table 3.6-1 for containment isolation valve closure times.
3.
All CIAS-actuated valves except MSIVs, MFIVs, and CCW Valves 2HV-6211, 2HV-6216, 2HV-6223 and 2HV-6236.
4a.
CCW noncritical loop isolation Valves 2HV-6212, 2HV-6213, 2HV-6218, and 2HV-6219 close.
4b.
Containment emergency cooler CCW isolation Valves 2HV-6366, 2HV-6367, 2HV-6368, 2HV-6369, 2HV-6370, 2HV-6371, 2HV-6372, and 2HV-6373 open.
5.
Response time includes instrumentation, logic, and isolation damper closure times only.
6.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
7.
Include HV4762 and HV4763 following implementation of DCP 195J.
Emergency diesel generator starting delay (10 sec.) and sequence loading delays for SIAS are included.
Emergency diesel generator starting delay (10 sec.) is included.
SAN ONOFRE - UNIT 2 3/4 3-30 AMENDMENT NO. 54
-~-
m TABLE 4.3-2 2
ENGINEERED SAFETY FEAluRE ACTUATION SYSTEM INSTRUMENTAION SURVEILLANCE RFQUIREMENTS S
S CilANNEL MODES FOR WillCil A
CilANNEL CllANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CilECK CAllBRATION TEST IS REQUIRED g
U 1.
SAFETY INJECTION (SIAS)
N a.
Manual (Trip Buttons)
N.A.
N.A.
R 1, 2, 3, 4 b.
Containment Pressure - High S
R H
1, 2, 3 c.
Pressurizer Pressure - Low S
R H
1, 2, 3 d.
Automatic Actuation Logic N.A.
N.A.
M(1)(3),SA(4) 1, 2, 3, 4 2.
CONTAINMENT SPRAY (CSAS) a.
Manual (Trip Buttons)
N.A.
N.A.
R 1, 2, 3 b.
Containment Pressure --
High - High S
R H
1, 2, 3 c.
Automatic Actuation Logic H.A.
N.A.
M(1)(3),SA(4) 1, 2, 3 4
3.
CONTAINMENT ISOLATION (CIAS) y a.
Manual CIAS (Trip Buttons)
N.A.
N.A.
R 1,2,3,4 g
b.
Manual SIAS (Trip Buttons)(5)
N.A.
H.A.
R 1,2,3,4 c.
Containment Pressure - High S
R H
1,2,3 d.
Automatic Actuation Logic N.A.
N.A.
M(1)(3), SA(4) 1, 2, 3, 4 4.
MAIN STEAM ISOLATION (MSIS) a.
Manual (Trip Buttons)
N.A.
N.A.
R 1,2,3 b.
Steam Generator Pressure - Low S R
M 1, 2, 3 c.
Automatic Actuation Logic N.A.
N.A.
M(1)(3), SA(4) 1, 2, 3 5.
RECIRCULATION (RAS) a.
Refueling Water Storage Tank - Low S
R H
1, 2, 3, 4 k
b.
Automatic Actuation Logic H.A.
N.A.
M(1)(3), SA(4) 1, 2, 3, 4 z
E 6.
CONTAINMENT COOLING (CCAS) z a.
Manual CCAS (Trip Buttons)
N.A.
N.A.
R 1,2,3,4 b.
Manual SIAS (Trip Buttons)
N.A.
N.A.
R 1, 2, 3, 4 c.
Automatic Actuation Logic N.A.
N.A.
M(1)(3),SA(4) 1, 2, 3, 4
,5 8
cn
TABLE 4.3-2 (Continued) z ENGIhEERED SAFETY FtATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS o
k CHANNEL MODES FOR WHICH g
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATIO_N TEST IS REQUIRED g
U 7.
LOSS OF POWER (LOV) m a.
4.16 kv Emergency Bus Undervoltage (Loss of Voltage and Degraded Voltage)
S R
R 1,2,3,4 8.
Manual (Trip Buttons)
N.A.
N.A.
R 1,2,3 b.
SG Level (A/B)-Low and AP (A/B) - High S
R M
1,2,3 c.
SG Level (A/B) - Low and Na
,g Pressure - Low Trip (A/B)
S R
H 1,2,3 d.
Automatic Actuation Logic N.A.
N.A.
M(1)(3), SA(4) 1, 2, 3 9.
CONTROL ROOM ISOLATION (CRIS) a.
Manual CRIS (Trip Buttons)
N.A.
N.A.
R N.A.
b.
Manual SIAS (Trip Buttons)
N.A.
N.A.
R N.A.
c.
Airborne Radiation i.
Particulate / Iodine S
R H
All
- 11. Gaseous S
R M
All d.
Actomatic Actuation Logic N.A.
N.A.
R(3)
All 10.
T0XIC GAS ISOLATION (TGIS) a.
Manual (Trip Buttons)
N.A.
N.A.
R N.A.
b.
Chlorine - High S
R M
All c.
Ammonia - High S
R M
All k
d.
Butane / Propane - High S
R M
All z
e.
Automatic Actuation Logic N. A.
N.A.
R (3)
All l
E 4
9 l
5 h
~
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETO NGS BASES Local Power Density-High (Continued)
The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines. These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the peak LPD Safety Limit.
CPC uncertainties related to peak LPD are the same types used for DNBR calculation.
Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
Departure from Nucleate Boiling Ratio (DNBR) - Low The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of anticipated operational occurrences.
The DNBR is calculated in the CPC i
utilizing the following information:
1 a.
Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; b.
Reactor Coolant System pressure from pressurizer pressure measurement; c.
Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements; d.
Radial peaking factors from the position measurement for the CEAs; e.
Reactor coolant mass flow rate from reactor coolant pump speed; f.
Core inlet temperature from reactor coolant cold leg temperature measurements.
The DNBR, the trip variable calculated by the CPC incorporates various uncertainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits.
These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core l
DNBR is sufficiently greater than 1.31 such that the decrease in actual core SAN ONOFRE-UNIT 2 B 2-5 AMENDMENT NO. 54
d SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR-Low (Continued)
DNBR after the trip will not result in a violation of the LNBR Safety Limit.
CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithic modelling uncertainties, and computer equipment processing uncer-tainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
The DNBR algorithm used in the CPC is valid only within the parametric envelope indicated below which is typically less restrictive than plant specific limiting conditions for operation.
Operation outside of these limits will re-sult in a CPC initiated trip.
a.
RCS Cold Leg Temperature-Low 1 495 F b.
RCS Cold Leg Temperature-High
< 580 F c.
Axial Shape Index-Positive
<+0.5 d.
Axial Shape Index-Negative 1-0.5 e.
Pressurizer Prec.sure-Low 1 1825 psia f.
Pressurizer Pressure-High
< 2375 psia g.
Integrated Radial Peaking Factor-Low 1 1.28 h.
Integrated Radial Peaking Factor-High
< 4.28 i.
Hot Leg Quality 0 (no net quality) l The DNBR Trip setpoint in CPC and COLSS is 1.31.
The values of the penalty factors BERR1 (CPC) and EPOL2 (COLSS) may be adjusted to implement requirements for tripping at other values of DNBR.
The following formula is used to adjust the CPC addressable constant BERR1:
BERR1 BERR1old [1 + ADNBR(%)* d B)
.00
=
new where:
BERR1 new required value of BERR1,
=
new BERR1old present implemented value of BERR1,
=
ADNBR(%)
percent increase in DNBR trip setpoint requirement,
=
d(% POL)/d(% DNBR)
The absolute value of the most adverse derivative
=
of percent POL with respect to percent DNBR as reported in CEN-184(S)-P.
Similarly, for the COLSS addressable constant EP0L2:
EPOL2 (1 + ADNBRf%)* d NB ) *0.01)*(1 + EP0L2old)-1.0
=
new where:
EPOL2 new required value of EPOL2,
=
new EPOL2 present implemented value of EPOL2,
=
old and the other terms are as previously defined.
SAN ONOFRE-UNIT 2 8 2-6 AMENDMENT N0. 54
3/4.2 POWER DISTRIBUTION LIMITS BASES i
3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating I
limit corresponding to the allowable peak linear heat rate.
Reactor operation at or below this calculated power level assures that the limits of 13.9 kw/ft are not exceeded.
The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator.
A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit.
This provides adequate margin to the linear heat rate operating limit for normal steady state opera-(
tion. Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power o'perating limit being exceeded.
In the event this occurs, COLSS alarms will be annunciated.
If the n
event which causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.
The COLSS calculation of the linear heat rate includes appropriate penalty factors which provide, with a 95/95 probability / confidence level, that the maximum linear heat rate calculated by COLSS is conservative with respect to the actual maximum linear heat rate existing in the core.
These penalty factors are determined from the uncer-tainties associated with planar radial peaking measurement, engineering design factors, axial densification, software algorithm modelling, computer processing, rod bow and, core power measurement.
l
'. Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNS and total core power are also monitored by the CPCs assuming minimum core power of 20% RATED THERMAL POWER. The 20% Rated i
Thermal Power threshold is due to the neutron flux det'.sctor system being 4
inaccurate below 20% core power.
Core noise level at low power is too large to obtain usable detector readings. Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-2 can be maintained by utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels.
The above listed uncertainty penalty factors plus those associated with startup test acceptance criteria are also included in the CPCs.
l l
SAN ONOFRE-UNIT 2 B 3/4 2-1 l
l
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 PLANAR RADIAL PEAKING FACTORS c
Limiting the values of the PLANAR RADIAL PEAKING FACTORS (F,) used in the COLSS and CPCs to valges equal to or greater than the measurH PLANAR RADIAL PEAKING FACTORS (F COLSSandtheCPCsremain*Va)lidprovides assurance that the limits calculated by Data from the incore detectors are used for determining the measured PLANAR RADIAL PEAKING FACTORS.
A minimum care power at 20% of RATED THERMAL POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS. The 20% Rated Thermal Power threshold is due to the neutron flux detector system being inaccurate below 20% core power.
Core noise level at low power is too large to obtain usable detector readings.
The periodic surveillance requirements for determining the measured PLANAR RADIAL PEAKING FACTORS provides assurance that the PLANAR RADIAL PEAKING FACTORS used in COLSS and the CPCs remain valid throughout the fuel cycle.
Determining the measured PLANAR RADIAL PEAKING FACTORS after each fuel loading prior to exceeding 70% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.
3/4.2.3 AZIMUTHAL POWER TILT - Tg j
The limitations on the AZIMUTHAL POWER TILT are provided to ensure that l
design safety margins are maintained.
An AZIMUTHAL POWER TILT greater than O.10 is not expected and if it should occur, operation is restricted to only j
those conditions required to identify the cause of the tilt. The tilt is normally calculated by COLSS. The surveillance requirements specified when l
COLSS is out of service provide an acceptable means of detecting the presence of a steady state tilt.
It is necessary to explicitly account for power asym-metries because the radial peaking factors used in the core power distribution calculations are based on an untilted power distribution.
This LCO does not apply below 20% rated thermal power for two reasons:
(1) The incore neutron detectors are inaccurate at low core power l u cis due to the poor signal-to-noise ratio which they experience. The resultant COLSS AZIMUTHAL POWER TILT is unreliable since COLSS uses the incore neutron detector signals to perform this calculation.
(2) The CPC's assume a minimum core power of 20% rated thermal power. When actual power is below this level the core is operating further from thermal 1
limits and the resultant CPC-calculated DNBR and LPD trips are highly conservative.
AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power at any core location in the presence of a tilt to the untilted power at the location is of the form:
P
/P
=1+T g cos (0 - e )
tilt untilt q
n where:
SAN ONOFRE - UNIT 2 B 3/4 2-2 AMENDMENT NO. 54
YEc o
A, UNITED STATES 8
p, NUCLEAR REGULATORY COMMISSION 5
nj W ASHINGTON, 0. C. 20555
%,...../
SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-362 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 43 License No. NPF-15 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment to the license for San Onofre Nuclear Generating Station, Unit 2 (the facility) filed by the Southern California Edison Company on behalf of itself and San Diego Gas and Electric Company, The City of Riverside and the City of Anaheim, California (licensees) dated April 2, and April 27, 1984 and March 18, and July 1, 1985, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as 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 regulations of the Comission; 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 Comission's regulations set forth in 10 CFR Chapter I; 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; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements i
have been satisfied.
. I l
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this amendment and Paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 43, are hereby incorporated in the license.
SCE shall operate the facility in accordance with the Technical f
Specifications and the Environmental Protection Plan.
3.
The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment.
In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during change over shall be minimized.
4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
Harry Roo, Senior Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-B l
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 9, 1986 l
September 9, 1986 ATTACHMENT TO LICENSE AMENDMENT NO. 43 FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 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 ir.dicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.
Amendment Page Overleaf Page 3/4 2-4 3/4 2-3 3/4 3-17 3/4 3-18 3/4 3-30 3/4 3-29 3/4 3-32 3/4 3-31 B 2-5 B 2-6 B 3/4 2-2 B 3/4 2-1
POWER DISTRIBUTION LIMITS 3/4.2.3 AZIMUTHAL POWER TILT - To LIMITING CONDITION FOR OPERATION 3.2.3 The AZIMUTHAL POWER TILT (T shall be less than or equal to the AZIMUTHAL POWER TILT Allowance use8)in the Core Pr APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.*
ACTION:
With the measured AZIMUTHAL POWER TILT determined to exceed the a.
AZIMUTHAL POWER TILT Allowance used in the CPCs but less than or equal to 0.10, within two hours either correct the power tilt or adjust the AZIMUTHAL POWER TILT Allowance used in the CPCs to greater than or equal to the measured value.
b.
With the measured AZIMUTHAL POWER TILT determined to exceed 0.10:
1.
Due to misalignment of either a part length or full length CEA, within 30 minutes verify that the Core Operating Limit
(
Supervisory System (COLSS), when COLSS is being used to monitor the core power distribution per Specifications 4.2.1 and 4.2.4, is detecting the CEA misalignment.
2.
Verify that the AZIMUTHAL POWER TILT is within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit or reduce THERMAL POWER to l
less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Linear Power Level - High trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the AZIMUTHAL POWER TILT is verified within its limit at least once par hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified acceptable at 95% or greater RATED THERMAL POWER.
i See Special Test Exception 3.10.2.
I SAN ONOFRE-UNIT 3 3/4 2-3
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3 The AZIMUTHAL POWER TILT shall be determined to be within the limit above 20% of RATED THERMAL POWER by:
a.
Continuously monitoring the tilt with COLSS when the COLSS is OPERABLE.
b.
Verifying at least once per 31 days, that the COLSS Azimuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT greater than the AZIMUTHAL POWER TILT Allowance used in the CPCs.
c.
Using the incore detectors at least once per 31 EFPD's to indepen-l dently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.
d.
Calculating the tilt at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the COLSS is inoperable.
SAN ONOFRE-UNIT 3 3/4 2-4 AMENDMENT NO. 43
t h
TABLE 3.3-3 (Continued)
~
Q ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION A
m E:
MINIMUM 3
TOTAL NO.
CHANNELS CHANNELS APPLICABLE
]
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 9.
CONTROL ROOM ISOLATION (CRIS) a.
Manual CRIS (Trip Buttons) 2 1
1 All 13*#
b.
Manual SIAS (Trip Buttons) 2 sets of 2/ unit 1 set of 2 2 sets of 2/ unit 1, 2, 3, 4 8
c.
Airborne Radiation un 1.
Particulate / Iodine 2
1 1
All 13*#
)
- 11. Gaseous 2
1 1
All 13*#
d.
Automatic Actuation u,
Logic 1/ train 1
1 All 13*#
10.
T0XIC GAS ISOLATION (TGIS) a.
Manual (Trip Buttons) 2 1
1 All 14*#, 15*#
b.
Chlorine - High 2
1 1
All 14*#, 15*#
c.
Ammonia - High 2
1 1
All 14*#, 15*#
d.
Butane / Propane - High 2
1 1
All 14*#, 15*#
e.
Automatic Actuation Logic 1/ train 1
1 All 14*#, 15*#
5
t
- s TABLE 3.3-3 (Continued) s:
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E
El N;
MINIMUM a
TOTAL NO.
CHANNELS CHANNELS APPLICABLE 2
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 11.
FUEL HANDLING ISOLATION (FHIS) a.
Manual (Trip Buttons) 2 1
1 16*#
b.
Airborne Radiation i.
Gaseous 2
1 1
16*#
ii. Particulate / Iodine 2
1 1
16*#
c.
Automatic Actuation Logic 1/ train 1
1 16*#
'J 12.
CONTAINMENT PURGE ISOLATION (CPIS) u, E
a.
Manual (Trip Buttons) 2 1
1 6
17b*#
b.
Airborne Radiation (3RT-7804-1 or 3RT-7807-2 i.
Gaseous 2
1 1
1,2,3,4 17a 6
17b*#
ii.
Particulate 2
1 1
1,2,3,4 17a 6
17b*#
iii. Iodine 2
1 1
6 17b*#
c.
Containment Area Radiation (Gamma) 2 1
1 1,2,3,4 17 (3RT-7856-1 or 3RT-7857-2) 6 17b*#
d.
Automatic Actuation P
Logic 1/ train 1
1 1,2,3,4 17 h
6 17b*#
E 5
8 n.
b
m
~
[
Table 3.3-5 (Continued)
INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 5.
Steam Generator Pressure - Low a.
MSIS (1) Main Steam Isolation (HV8204, HV8205) 6.9 (2) Main Feedwater Isolation (HV4048, HV4052) 10.9 l
(3) Steam, Blowdown and Sample Isolation 20.9 (HV8419, HV8421)
(HV4053,HV4054,HV4057,HV4058)
(4) Auxiliary Feedwater Isolation (NOTE 7) 40.9 (HV4705, HV4713, HV4730, HV4731)
(HV4706,HV4712,HV4714,HV4715) 6.
Refueling Water Storage Tank - Low a.
RAS (1) Containment Sump Valves Open 50.7*
7.
4.16 kV Emergency Bus Undervoltage a.
LOV (loss of voltage and degraded voltage)
Figure 3.3-1 8.
Steam Generator Level - Low (and No
(
Pressure-Low Trip) a.
EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**
(2) Auxiliary Feedwater (Steam /DC train) 42.7 (Note 6) i 9.
Steam Generator Level - Low (and P - High) a.
EFAS (1) Auxiliary Feedwater (AC trains) 52.7*/52.7**
]
(2) Auxiliary Feedwater (Steam /DC train) 42.7 (Note 6) l
- 10. Control Room Ventilation Airborne Radiation a.
CRIS (1) Control Room Ventilation - Emergency Mode Not npplicable
- 11. C^ontrol Room Toxic Gas (Chlorine) a.
TGIS (1) Control Room Ventilation - Isolation Mode 16 (NOTE 5)
- 12. Control Room Toxic Gas (Ammonia) a.
TGIS (1) Control Room Ventilation - Isolation Mode 36 (NOTE 5)
SAN ONOFRE - UNIT 3 3/4 3-29 AMENDMENT NO. 35
Table 3.3-5 (Continued)
INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC)
- 13. Control Room Toxic Gas (Butane / Propane)
TGIS Control Room Ventilation -
Isolation Mode 36 (NOTE 5) 14.
Fuel Handling Building Airborne Radiation FHIS Fuel Handling Building Post-Accident Cleanup Filter System Not Applicable 15.
Containment Airborne Radiation CPIS Containment Purge Isolation 2 (NOTE 2) 16.
Containment Area Radiation CPIS Containment Purge Isolation 2 (NOTE 2)
NOTES:
1.
Response times include movement of valves and attainment of pump or blower discharge pressure as applicable.
2.
Response time includes emergency diesel generator starting delay (applicable to AC cotor operated valves other than containment purge valves), instrumentation and logic response only.
Refer to Table 3.6-1 for containment isolation valve closure times.
3.
All CIAS-Actuated valves except MSIVs and MFIVs and CCW valves 3HV-6211, 3HV-6216, 3HV-6223 and 3HV-6236.
4a.
CCW non-critical loop isolation valves 3HV-6212, 3HV-6213, 3HV-6218 and 3HV-6219.
4b.
Containment emergency cooler CCW isolation valves 3HV-6366, 3HV-6367, 3HV-6368, 3HV-6369, 3HV-6370, 3HV-6371, 3HV-6372 and 3HV-6373 open.
5.
Response time includes instrumentation, logic, and isolation damper closure times only.
6.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
7.
Include HV4762 and HV4763 following implementation of DCP 195J.
l Emergency diesel generator starting delay (10 seconds) and sequence loading delays for SIAS are included.
Emergency diesel generator starting delay (10 seconds) is included.
1 SAN ONOFRE - UNIT 3 3/4 3-30 AMENDMENT NO. 43 l.
l 1
- y/
.r vs E
TABLE 4.3-2 5
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAION SURVEILLANCE c3 2?
9?
CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE Si FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED
- 2; 1.
SAFETY INJECTION (SIAS) u, a.
Manual (Trip Buttons)
N.A.
N.A.
R 1, 2, 3, 4 b.
Containment Pressure - High S
R M
1,2,3 c.
Pressurizer Pressure - Low 5
R H
1,2,3 d.
Automatic Actuation Logic N A.
N. A.
M(1)(3), SA(4) 1, 2, 3, 4 2.
CONTAINMENT SPRAY (CSAS) a.
Manual (Trip Buttons)
N.A.
N.A.
R 1,2,3 b.
Containment Pressure --
High - High S
R M
1,2,3 Automatic Actuation Logic N.A.
N.A.
M(1)(3),SA(4) 1, 2, 3 u,
c.
3:
3.
CONTAINMENT ISOLATION (CIAS) c.
J, a.
Manual CIAS (Trip Buttons)
N.A.
N.A.
R 1,2,3,4 b.
Manual SIAS (Trip Buttons)(5)
N.A.
N. A.
R 1,2,3,4 Containment Pressure - High S
R M
1,2,3 c.
d.
Automatic Actuation Logic N. A.
N.A.
M(1)(3), SA(4) 1, 2, 3, 4 4.
MAIN STEAM ISCLATION (MSIS)
Manual (Trip Buttons)
N.A.
N.A.
R 1,2,3 a.
b.
Steam Generator Pressure - Low S R
M 1,2,3 Automatic Actuation Logic N. A.
I,*. A.
M(1)(3), SA(4) 1, 2, 3 c.
5.
RECIRCULATION (RAS)
Refueling Water Storage a.
Tank - Low S
R M
1,2,3,4 b.
Automatic Actuation Logic N.A.
N.A.
M(1)(3), SA(4) 1, 2, 3, 4 6.
CONTAINMENT COOLING (CCAS) a.
Manual CCAS (Trip Buttons)
N.A.
N.A.
R 1,2,3,4 b.
Manual SIAS (Trip Buttons)
N.A.
N.A.
R 1,2,3,4
)
c.
Automatic Actuation Logic N.A.
N.A.
M(1)(3), SA(4) 1, 2, 3, 4
a TABLE 4.3-2 (Continued) z ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
CHANNEL MODES FOR WHICH a
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED W
7.
LOSS OF POWER (LOV) a.
4.16 kV Emergency Bus Undervoltage (Loss of Voltage and Degraded Voltage)
S R
R 1,2,3,4 8.
Manual (Trip Buttons)
N.A.
N.A.
R 1,2,3 b.
SG Level (A/B)-Low and AP (A/B) - High S
R M
1,2,3 w}
c.
SG Level (A/B) - Low and No Pressure - Low Trip (A/B)
S R
M 1,2,3 6
d.
Automatic Actuation Logic N.A.
N.A.
M(1)(3), SA(4) 1, 2, 3 m
9.
CONTROL ROOM ISOLATION (CRIS) a.
Manual CRIS (Trip Buttons)
N.A.
N.A.
R N.A.
b.
Manual SIAS (Trip Buttons)
N.A.
N.A.
R N.A.
c.
Airborne Radiation 1.
Particulate / Iodine S
R M
All
- 11. Gaseous S
R M
All d.
Automatic Actuation Logic N.A.
N.A.
R(3)
All 10.
T0XIC GAS ISOLATION (TGIS) a.
Manual (Trip Buttons)
N.A.
N.A.
R N.A.
b.
Chlorine - High S
R M
All k
c.
Ammonia - High S
R M
All g
d.
Butane / Propane - High S
R M
All g
e.
Automatic Actuation Logic N.A.
N.A.
R (3)
All l
4 5
t
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS j
i BASES l
Local Power Density-High (Continued)
The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines.
These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the peak LPD Safety Limit.
CPC uncertainties related to peak LPD are the same types used for DNBR calculation.
Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
(
l Departure from Nucleate Boiling Ratio (DNBR) - Low I
i The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of anticipated operational occurrences.
The DNBR is calculated in the CPC utilizing the following information:
a.
Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; b.
Reactor Coolant System pressure from pressurizer pressure measurement; c.
Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements; d.
Radial peaking factors from the position measurement for the CEAs; j
e.
Reactor coolant mass flow rate from reactor coolant pump speed; f.
Core inlet temperature from reactor coolant cold leg temperature measurements.
The DNBR, the trip variable calculated by the CPC incorporates various uncertainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits.
These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core DNBR is sufficiently greater than 1.31 such that the decrease in actual core l
l SAN ONOFRE - UNIT 3 8 2-5 AMENDMENT NO. 43
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR-Low (Continued)
DNBR after the trip will not result in a violation of the DNBR Safety Limit.
CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncer-1 tainties.
Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
The DNBR algorithm used in the CPC is valid only within the parametric envelope indicated below which is typically less restrictive than plant specific limiting conditions for operation.
Operation outside of these limits will re-sult in a CPC initiated trip.
a.
RCS Cold Leg Temperature-Low 1 495*F b.
RCS Cold Leg Temperature-High
< 580 F c.
Axial Shape Index-Positive
<+0.5 d.
Axial Shape Index-Negative 1-0.5 e.
Pressurizer Pressure-Low 1 1825 psia f.
Pressurizer Pressure-High
< 2375 psia g.
Integrated Radial Peaking Factor-Low 1 1.28 h.
Integrated Radial Peaking Factor-High
< 4.28 i.
Hot Leg Quality 0 (no net quality)
The DNBR Trip setpoint in CPC and COLSS is 1.31.
The values of the penalty factors BERR1 (CPC) and EPOL2 (COLSS) may be adjusted to implement requirements for tripping at other values of DNBR. The following formula is used to adjust the CPC addressable constant BERR1:
BERR1 BERR1old [1 + ADNBR(%)* d 08)
.00
=
new where:
BERR1 new required value of BERR1,
=
new BERR1 present implemented value of BERR1,
=
old ADNBR(%)
percent increase in DNBR trip setpoint requirement,
=
d(% POL)/d(% DNBR)
The absolute value of the most adverse derivative
=
of percent POL with respect to percent DNBR as reported in CEN-184(S)-P.
Similarly, for the COLSS addressable constant EPOL2:
EPOL2 (1 + ADNBR(%)* d NB ) *0.01)*(1 + EPOL201d)-1.0
=
new where:
EPOL2 new required value of EPOL2,
=
new EPOL2 present implemented value of EPOL2,
=
old and the other terms are as previously defined.
SAN ONOFRE - UNIT 3 8 2-6 AMENDMENT NO. 43
r 3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate.
Reactor operation at or below this calculated power level assures that the limits of 13.9 kw/ft are not exceeded.
The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steady state opera-tion. Normal reactor power transients or equipment failures which do not
(
require a reactor trip may result in this core power operating limit being f
exceeded.
In the event this occurs,.COLSS alarms will be annunciated.
If the event which causes the COLSS limit to. be exceeded results in conditions which approach the core safety limits, a reactor trip will be inP by the Reactor Protective Instrumentation. The COLSS calculation of the linear heat rate includes appropriate penalty factors which provide, with a 95/95 probability / confidence level, that the maximum linear heat rate calculated by COLSS is conservative with respect to the actual maximum linear heat rate
(
existing in the core. These penalty factors are determined from the uncer-l tainties associated with planar radial peaking measurement, engineering design I
factors, axial densification rod bow and core power measur,ement. software algorithm modelling, computer processing, Parameters required to maintain the operating limit power level based on linear heat rate, margin to DN8 and total core power are aisc snaitored by the CPCs assuming minimum core power of 20% RATED THERMAL POWER.
The 20% Rated Thermal Power threshold is due 4 the neutron flux detector system being inaccurate below 20% core power. Core noise level at low power is too large to obtain usable detector readings.
Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-2 can be maintained by utilizing a predetennined local power density margin and a total core power limit in the CPC trip channels.
The above listed uncertainty penalty factors plus those associated with startup test acceptance criteria are also included
{
in the CPCs.
SAN ONOFRE-UNIT 3 8 3/4 2-1
1 i
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 PLANAR RADIAL PEAKING FACTORS c
Limiting the values of the PLANAR RADIAL PEAKING FACTORS (F,) used in the COLSS and CPCs to valges equal to or greater than the measurH PLANAR RADIAL PEAKING FACTORS (F COLSSandtheCPCsremain*Va) lid;provides assurance that the limits calculated by Data from the incore detectors are used for determining the measured PLANAR RADIAL PEAKING FACTORS.
A minimum core power at 20% of RATED THERMAL POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS.
The 20% Rated Thermal Power threshold is due to the neutron
' flux detector system being inaccurate below 20% core power.
Core noise level at low power is too large to obtain usable detector readings. The periodic surveillance requirements for determining the measured PLANAR RADIAL PEAKING FACTORS provides assurance that the PLANAR RADIAL PEAKING FACTORS used in COLSS and the CPCs remain valid throughout the fuel cycle.
Determining the measured PLANAR RADIAL PEAKING FACTORS after each fuel loading prior to exceeding 70% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.
3/4.2.3 AZIMUTHAL POWER TILT - Tq The limitations on the AZIMUTHAL POWER TILT are provided to ensure that design safety margins a're maintained.
An AZIMUTHAL POWER TILT greater than 0.10 is not expected and if it should occur, operation is restricted to only those conditions required to identify the cause of the tilt.
The tilt is normally calculated by COLSS.
The surveillance requirements specified when l
COLSS is out of service provide an acceptable means of detecting the presence of a steady state tilt.
It is necessary to explicitly account for power asym-metries because the radial peaking factors used in the core power distribution calculations are based on an untilted power distribution.
]
This LCO does not apply below 20% rated thermal power for two reasons:
(1) The incore neutron detectors are inaccurate at low core power levels due j
to the poor signal-to-noise ratio which they experience.
The resultant COLSS AZIMUTHAL POWER TILT is unreliable since COLSS uses the incore neutron detector signals to perform this calculation.
(2) The CPC's assume a minimum core power of 20% rated thermal power. When actual power is below this level the core is operating further from thermal limits and the resultant CPC-calculated DNBR and LPD trips are highly conservative.
1 AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power at any core location in the presence of a tilt to the untilted power at the location is of the form:
P
/P
=1+T g cos (0 - s )
tilt untilt q
o 3
where:
SAN ONOFRE - UNIT 3 8 3/4 2-2 AMENDMENT NO. 43 1
=.
m-
_