ML20095B602
| ML20095B602 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/28/1992 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20095B605 | List: |
| References | |
| NUDOCS 9204230025 | |
| Download: ML20095B602 (26) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION j
- k~1%/,/
WASmNGTON, n C. 20553 g
s.,
SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET N0. 50-361 SAN ON0FRE NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.101 License No. NPF-10 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Southern California Edison Company, et al. (SCE or the licensee) dated August 30, 1991, complies with the standards and requirements of the Atomic Entrgy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, 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; l
D.
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 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.
l 9204230025 920228 PDR ADOCK 05000361 P
. 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license 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. 101 are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuar.ce and must be fully implemented no later than 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ur/t1A 09 9
Theodore R. Quay, Director Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 28, 1992 l
l l
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.101 TO FACILITY OPERATING LICENSE NO. NPF-10 DOCKET NO. 50-361
-Revi:e Appendix A Technic 61 Specifications by removing the pages identified below and inserting tne enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-12a 3/4 3-12a 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 B 3/4 3-1 B 3/4 3-1 l
i
I
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E TABLE 4.3-1 E
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLAilCE REQUIREMENTS E
~*
CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 1.
Manual Reactor Trip N.A.
N.A.
1, 2, 3*, 4*, 5*
2.
Linear Power Level - High S
D(2,4),M(3,4), Q 1, 2 Q(4),#(4) 3.
Logarithmic Power Level - High S
- (4)
Q and S/U(1) 1,2,3,4,5 4.
Pressurizer Pressure - High S
Q 1, 2 5.
Pressurizer Pressure - Low 5
Q 1, 2 6.
Containmes, Pressure - High S
Q 1, 2 7.
Steam Generator Pressure - Law S
Q 1, 2 8.
Steam Generator Level - Low S
Q 1, 2 9.
Local Power Density - High 5
D(2,4),
Q,#(6) 1, 2
- (4,5) 10.
DNBR - Low S
S(7),D(2,4),
Q,#(6) 1, 2 M(8), #(4,5)
E 11.
Steam Generator Level - High S
Q 1, 2 9E 12.
Logic N.A.
N.A.
Q 1, 2, 3*, 4*, 5*
5
[$
TABLE 4.3-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS T
C5 CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED
- 13. : Reactor Trip Breakers H.A.
N.A.
M,(12) 1, 2, 3*, 4*, 5*
- 14., Core Protection Calculators S
0(2,4),S(7)
Q(11),#(6) 1, 2
- (4,5),M(8) 15.
CEA Calculators S
Q,#(6) 1, 2 16.
Reactor Coolant Flow-Low S
Q 1, 2 17.
Seismic-High S
Q 1, 2 U
18.
Loss of Load 5
N.A.
Q 1 (9) 3E I
i M
"i 5a 3
t
-..~.
TABLE 4.3-1 (Continued)
TABLE NOTATION (11) -
The quarterly CHANNEL FUNCTIONAL TEST shall include verification that the correct values of addressable constants are installed in each OPERABLE CPC.
(12) -
At least once per 18 months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage and shunt trips.
l l
l SAN ONOFRE-UNIT 2 3/4 3-12a AMENDMENT'NO. 101
w I
TABLE 4.3-2 m
E ENGINEERED SAFETY FEATURE ACTUATION SYSTLM INSTRUMENTATION SURVEILLANCE REQUIREMENTS o5g CHANNEL MODES FOR WHICH
'P CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g
FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED Z
1.
SAFETY INJECTION (SIAS) g a.
Manual (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 b.
Containment Pressure - High S
(6)
Q 1,2,3 c.
Pressurizer Pressure - Low S
(6)
Q 1, 2, 3, d.
Automatic Actuation Logic N.A.
N. A.
Q(3),SA(4) 1, 2, 3, 4 2.
CONTAINMENT SPRAY (CSAS) a.
Manual (Trip Buttons)
N. A.
N.A.
(6) 1, 2, 3 b.
Containment Pressure --
High - High 5
(6)
Q 1,2,3 c.
Automatic Actuation Logic N.A.
N.A.
Q(3), SA(4) 1, 2, 3 ms
[
3.
CONTAINMENT ISOLATION (CIAS)
J, a.
Manual CIAS (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 b.
Manual SIAS (Trip Buttons)(5)
N. A.
N.A.
(6) 1, 2, 3, 4 c.
Containment Pressure - High S
(6)
Q 1, 2, 3 d.
Automatic Actuation Logic N. A.
N.A.
Q(3),SA(4) 1, 2, 3, 4 4.
MAIN STEAM ISOLATION (MSIS) a.
Manual (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3 b.
Steam Generator Pressure - Low 5 (6)
Q 1, 2, 3 j
t c.
Automatic Actuation Logic N.A.
N.A.
Q(3),SA(4) 1, 2, 3 S.
RECIRCULATION (RAS) a.
Refueling Water Storage
,g Tank - Low 5
R Q
1,2,3,4 g
b.
Automatic Actuation Logic N.A.
N.A.
Q(3),SAb) 1, 2, 3, 4 x
E CONTAINMENT COOLING (CCAS)
[
6.
4 a.
Manual CCAS (Trin Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4
.o b.
Manual SIAS (Tr p Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 l
g c.
Automatic Actuation Logic N.A.
N. A.
Q(3),SA(4) 1, 2, 3, 4 l
w 1
\\
I TSBLE 4.3-2 (Continuad)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTA l
[
f CHANNEL PODES FOR Mi1CH
{
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE z
IS REQUIRED _
CHECK CALIBRATION TES1 f
h FUNCTIONAL UNIT I
=
LDSS OF PGGR (LCV) 7.
4.16 kv Emergency Bus a
Undervoltage (Loss of 1, 2, 3, 4 Voltage and Degraded 5
(6)
(6)
Voltage)
N.A.
N.A.
(6) 1, 2, ',
8.
Manual (Trip Buttons) 5 (6)
Q 1, 2, 3 I
a.
b.
SG Level (A/B)-Low and AP (A/B) - High 1, 2, 3 SG Level (A/B) - Low and No 5
(6)
Q N.A.
N.A.
Q(3),SA(4) 1, 2, 3 c.
R Pressure - Lew Trip' (A/B) d.
Autoteatic Actuation Logic Y
N.A.
CCNTROL ROOM ISOLATION (CRIS)
N.A.
N.A.
R O
9.
N.A.
N.A.
R N.A.
Manual CRIS (Trip Buttons) a.
Manual SIAS (Trip Buttons) b.
All Airborne Radiation 5
R M
All c.
i.
Particulate / Iodine S
R M
All ii. Gaseous N.A.
N.A.
R(3) d.
Automatic Actuation logic N.A.
- 10. T0XIC GAS ISOLATION (TGIS)
N.A.
N.A.
R All Manual (Trip Buttons) 5 R
M All a.
b.
Chlorine - High 3
R H
All Amonia - High S
R H
N.A.
N.A.
R (3)
All c.
M d.
Butane / Propane - High Automatic Actuation Logir E
e.
E2 e
E E
I
TABLE 4.3-2 (Continued) m "x
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTGUMENTATION SURVEILLANCE REQUIREMENTS o*o CHANNEL 10E5 FOR WHICH h
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g
FUNCTIONAL UNIT CHECK CALIBRATION TEST 15 REQUIRED 11.
FUEL HANDLING ISOLATION (FHIS) m a.
Manual (Trip Buttons)
N.A.
N R
N.A.
b.
Airterne Radiation
- i. Gaseous 5
R M
c.
Automatic Actuation Logic N.A.
N.A.
R(3) 12.
CONTAINMENT PtiRGE ISOLATI0H (CPIS) a.
Manual (Trip Buttons)
N.A.
N.A.
(5)
N.A.
b.
Airborne Radiation i.
Gasecus 5
(6)
M 1,2,3,4,6 ii.. Particulate W
(6)
M 1,2,3,4,6 R
iii. Iodine W
(6)
M 6
c.
Conta;nment Area Radiation i
y (Gama) 5 (6)
M 1,2,3,4,6 l
g 4.
Automatic Actuatian Logic N.A.
N.A.
(3),(6) 1,2,3,4,6 TABLE NOTATION l
(1) Deleted.
1 (2) Deleted.
(3) Testing of Automatic Actuation sogic shall include energization/de energization of each initiation relay and verification of tne OPERABILITY of each initiation relay.
(4) A subgroup relay test shall be performed whicle shall include the energization/de-energization cf each subgroup relay and verification of the OPERABILITY of each s::bgroup relay.
Relays exempt from testing during plant operation shall be limited to only those relays associated with plant equipment which
,x9 cannot be operated during plant operation.
Relays not testable during plant operation shall be tested E
during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless tested during the previous 6 months.
"i (5) Act.eated equipment only; does not result in CIAS.
,5 (6) At least once per refueling interval.
With irradiated fuel in the storage pool.
t 1NSTRUMENTAT10N 3/4.3.3 M3NITORING INSTRUMENTATION RADIAT)0N H3N170RIN3 INSTRUMENTATION LIMIT 1N3 CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3 6 i
shall be OPERABLE with their alarm / trip setpoints within the specified Itaits.
APPLICA?!LITY: As shown in Table 3.3-6.
ACTION:
With a radir. tion monitoring channel alarm / trip setpoint exceeding the a.
value shoin in Table 3.3 6, adjust the setpoint to within the liett 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 inoperabla.
b.
With one or core radiation monitoring channels inoptrable, take the ACTION shown in fable 3.3 6.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
c.
SURVEILLAN:E REQUIREMENTS 4.3.3.1 Ea:h radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the perfornance of the CHANNEL CHECK CHANNEL CALIBRATION and CHANNE FUNCTIONAL TECT operations for the HDDES and at the frequencies shown in Table 4.3 3.
I SAN ON0'RE-UNIT 2 3/4 3-34 AM'.NDHENT NO. 83 r _,,
1
.'L /4. 3 INSTRUMENTATION
, BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the reactor protective and Engineered Safety Features Actuation System instrumentation and bypasses ensure that 1) the associated Engineered Safety Features Actuation System action and/or reactor trip will be initietcd when the parameter monitored by ecch channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained,
- 3) sufficient redundu.cy is mair,ttined tc 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 sy.+ cms is required to provide the overall reliability, redundancy and diversity assumed available in the ficility design for the protection and mitigation of accident and transient conditions. The irtcgrated operation of each of these systems is consistent with the assun.ptions used in the accident analyses.
When a protection channel nf a given process variable becomes inoperable, the inoperable channel may be p. aced in bypass until the next Onsite Review Committee meeting at which time the Onsite Review Comittee will review and document-their judgment concerning prolonged operation in bypass, channel trip, and/cr repair. The goal shall be to return the inoperable channel to service as soon as practicable but in no case later than during the next COLD SHUTDOWN. This approach to bypass / trip in four channel protection systems is consistent with the applica'le criteria of IEEE Standards 279, 323, 344 and 384.
The Core Protection Cat slator (CPC) addressable constants are provided to allow calibration of the LPC system to more accurate indications of power level, RCS flow rate, &xial flux shape, radial peaking factors and CEA deviation penalties. Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inadvertent nis'oading of addressable constants into the CPCs is unlikely.
The redundancy and design of the Control Element Assembly Calculatoc" (CEAC) provides reactor protection in the event one or both CEAC's becomes in-operable.
If one CEAC is in test or inoperable, verification of CEAC position is performed at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If the second CEAC fails, the CPC's will use DNBR and LPD penalty factors, which restrict reactor operation to some j
maximum fraction of RATED THERMAL POWER.
If this maximum fraction is exceeded l
a reactor trip will occur.
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum I
frequencies are sufficient to demonstrate this capability. The quarterly l
frequency for the CHANNEL FUNCTIONAL TESTS for these systems is based on the analyses presented in the NRC approved topical report, CEN-327, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented.
The measurement of response time at the specified frequencies provides assurance that the reactor protective and ESF actuation associated with each l
channel is ccnpleted within the time limit assumed in the accident analyses.
1 SAN ON0FRE - UNIT 2 B 3/4 3-1 AMENDMENT NO.101
1 r
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
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.
Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement sensors with certified response tires.
l SAN ONOFRE - UNIT 2 B 3/4 3-la AMENDMENT NO 47 l
. ~.
,.[s>*htGy,'g UNITE D STATES
',g NUCLE AR REGULATORY COMMISSION y
,m W A5HtNOTON, D. C. 20L5$
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SOUTHERN CALIFORNIA EDISON COMPANY 3AN DIEGO GAS AND ELECTRIC COMPANY i
THE CITY OF RIVERSIDE, CAllFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-302 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 3 AMENDMENT TO FACILITY Or%C"tG LRENSE Amendment No. 90 Lit.onse No. NPF-15 1.
The Nuclear Regulatory Commission (the Commissim) has found that-A.
The applicaticn for amendment by Southern California Edison Company, et al. (SCE or the licensee) dated August 30, 1991, complies with the standards and requirenents of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, 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 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.
2 2.
Accordingly, the liconse is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPT-16 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. 90, are hereby incorporated in the license. Southern California Edison Comaany shall operate the facility in accordance with the Tec1nical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and must be fully implemented no later than 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION cde h #2 ucp Theodore R. Quay, Director Project Directorate V Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 28, 1992
ATTACHMENT _TO LICENSE AMENDMENT AMENDMENT NO. 90 TO FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 4
Revise Appendix A Technical Specifications by removing the pages identified belos and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 3-10 3/4 3-10 j
3/4 3-11 3/4 3 11 3/4 3-12a 3/4 3-12a 3/4 3-31 3/4 3 31
- 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 D 3/4 3-1 B 3/4 3-1 p
i a-
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- e. w,-
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,..-m.
,---,,-,vw,
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h TABLE 4.3-1 t
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REGUIREMENTS t
i m.
l c
CHANNEL MODES FOR WHICH 5
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE
[
FUNCTIONAL UNIT CHECK CALIBRATION TEST 15 REQUIRED 1.
Manual Reactor Trip N. A.
N.A.
1, 2, 3*, 4*, 5*
l 2.
Linear Power Level - High S
D(2,4),M(3,4), Q 1, 2 Q(4),#(4) 3.
Logarithmic Power Level - High 5
- (4)
Q and S/U(1) 1,2,3,4,5 i
4.
Pressurizer Pressure - High 5
Q 1, 2 5.
Pressurizer Pressure - Low S
Q 1, 2 i
6.
Containment Pressure - High 5
Q 1, 2
[
7.
Steam Generator Pressure - Low 5
Q 1, 2 I
8.
Steam Generator Level - Low S
Q 1, 2 r
9.
Local Power Density - High S
D(2,4),
Q,f(6) 1, 2 f(4,5) i h
- 10. DNBR - Low S
S(7),D(2,4),
Q,#(6) 1, 2 j
M(8),#(4,5) i 11.
Steam Generator Level - High S
Q 1, 2 i
t i
E 12.
't togic N.A.
N.A.
Q 1, 2, 2*, 4*, 5*
5 5
i 1
g 1
m--. -
I y
tat!E4.3-1(Continued) i E
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS A
'?
1 c-CHANNEL MODES FOR WICH 5
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE
[
FUNCTIONAL UNIT CHECK _
CALIBRATION TEST IS REQUIRED 13.
Reactor Trip Breakers N.A.
N.A.
M,(12) 1, 2, 3*, 4*, 5*
1 14.
Core Protection Calculators S
D(2,4),S(7), Q(11),#(6) 1, 2
- (4,5),M(8) 15.
CEA Calculators S
Q,#(6) 1, 2 16.
Reactor Coolant Flow-Low S
Q 1, 2 y 17. Seismic-High S
Q 1, 2 18.
Loss of load 5
N.A.
Q 1 (9) 2E=
E 2.o" i
TABLE 4.3-1 (Continued) i
_ TABLE NDTATION With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.
At least once per Refueling Interval, l
(1)
Each startup or when required with the reactor trip breakers c1cled and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.
(2)
Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15%
of RATED THERKAL POWER; adjust the Lineer Power Level signals end ine CPC addressable constant multipliers to make the CPC delta T j
power and CPC nuclear power calculations agree with the calorimetric calculation if absolute difference is greater than 2%.
During PHYSICS TESTS, these daily calibrations may be suspended provided these cali-brations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.
(3)
Above 15% of RATED THERMAL POWER, verify that the linear power subchannel gains of the encore detectors are consistent with the values used to estab-lish the shape annealing matrix elements in the Core Protection Calculators.
(4)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5)
Af ter each fuel loading and prior to exceeding 70% of RATED THERHAL POWER, the incere detectors shall be used to determine the shape annealing matrir, elements and the Core Protection Calculators shall use these elements.
(6)
This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions.
(7)
Above 70% of RATED THERMAL POVER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pres-sure instrumentation (conservatively compensated for measurement uncertain-l ties) or by calorimetric calculations (conservatively compensated for measurement uncertainties) and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than l
or equal to the actual flow rate.
The flow measurement uncertainty may be j
included in the BERR1 ters in the CPC and is equal to or greater than 4%.
(8)
Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations (conservatively compensated for measurement uncertainties).
(
(9)
Above 55% of RATED THERMAL POWER.
(10)-
Deleted.
l SAN ONOFRE
'JHIT 3 3/4 3 12 AMENDMENT NO. 7F JUN 8 1930
TABLE 4.3-1 (Continued)
TABLE NOTATION (11) -
The quarterly CHANNEL FUNCTIONAL TEST shall include verification that l
the correct values of addrestable constants are installed in each CPERABLE CPC.
(12)-
Atleastonceper18monthsandfollowingmaintenanceoradjustmentof the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage and shunt trips.
l l
l.
l l
i l-SAN ONOFRE - UNIT 3 3/4 3-12a AMENDMENT NO.90 l
l w,o es...,,
.-.-s.e
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w-.-,--r_,,,,-e mm.,ry,w.
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wg-TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS A
m E-CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 1.
SAFETY INJECTION (SIAS) a.
Manual (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 b.
Containment Pressure - High S
(6)
Q 1,2,3 i
c.
Pressurizer Pressure - Low S
(6)
Q 1,2,3 d.
Automatic Actuation Logic N.A.
N.A.
Q(1)(3),SA(4) 1, 2, 3, 4 2.
CONTAINMENT SPRAY (CSAS) a.
Manual (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3 l
b.
Containment Pressure --
High - High 5
(6)
Q 1, 2, 3
{
c.
Autor.atic Actuation Logic N.A.
N.A.
Q(1)(3),SA(4) 1, 2, 3 i
T 3.
CONTAINMENT ISOLATION (CIAS) a.
Manual CIAS (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 b.
Manual SIAS (Trip Buttons)(S)
N.A.
N.A.
(6) 1, 2, 3, 4 c.
Containment Pressure - High S
(6)
Q 1, 2, 3 g
d.
Automatic Actuation Logic N.A.
N.A.
Q(1)(3),SA(4) 1, 2, 3, 4 i
4.
MAIN STEAM ISOLATION (MSIS) a.
Manual (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3 b.
Steam Generator Pressure - Low S (6)
Q 1, 2, 3 c.
Automatic Actuation Logic N.A.
N.A.
Q(1)(3), SA(4) 1, 2, 3 S.
RECIRCULATION (RAS) a.
Refueling Water Storage E
Tank - Low S
R Q
1, 2, 3, 4 y
b.
Automatic Actuation Logic N.A.
N.A.
Q(1)(3),SA(4) 1, 2, 3, 4 E
6.
CONTAINMENT COOLING (CCAS) a.
Manual CCAS (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 5
b.
Manual SIAS (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3, 4 g
1 c.
Automatic Actuation Logic N.A.
N. A.
Q(1)(3),SA(4) 1, 2, 3, 4 l
o t
E TABLE 4.3-2 (Continued) z
,E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 2
m CHANNEL MODES FOR WHICH c'-
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 7.
LOSS OF POWER (LOV) a.
4.16 kV Emergency Bus Undervoltage (Loss of Voltage and Degraded j
Voltage) 5 (6)
(6) 1, 2, 3, 4 8.
Manual (Trip Buttons)
N.A.
N.A.
(6) 1, 2, 3 b.
SG Level (A/B)-Low and AP (A/B) - High S
(6)
Q 1, 2, 3 l
R c.
SG Level (A/B) - Low and No t
Pressure - Low Trip (A/B) 5 (6)
Q 1,2,3 d.
Automatic Actuation Logic N.A.
N.A.
Q(3),SA(4)
I, 2, 3 4
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 i.
Particulate / Iodine S
R H
All ii. 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 5
R M
All R
c.
Ammonia - High 5
R M
All G
d.
Butane / Propane - High 5
R M
All M
e.
Automatic Actuation Logic N.A.
N.A.
R (3)
All E
O
'g 4
l u,
i g
TABLE 4.3-2 (Continued) 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS "w
T CHANNEL MODES FOR WHICH E
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE t
Z FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 11.
FUEL HANDLING ISOLATION (FHIS) a.
Manual (Trip Buttons)
N.A.
N.A.
R N.A.
b.
Airborne Radiation
- i. Gaseous S
R M
i c.
Automatic Actuation Logic N.A.
N.A.
R(3) i 12.
CONTAINMENT PURGE ISOLATION (CPIS) a.
Manual (Trip Buttons)
N.A.
N.A.
(6)
N. A.
b.
Airborne Radiation i.
Gaseous 5
(6)
M 1,2,3,4,6
(
,}
ii.
Particulate W
(6)
M 1,2,3,4,6 i
iii. Iodine W
(6)
M 6
?
wa c.
Containment Area Radiation (Gamma)
S (6',
M 1,3,3,4,6 w
d.
Automatic Actuation Logic N.A.
N.A.
(3), (6) 1,2,3,4,6 L
TABLE NOTATION (1) Deleted.
l (2) Deleted.
(3) Testing of Automatic Actuation Logic shall include energization/de-energization of each initiation relay and verification of the OPERABILITY of each initiation relay.
g (4) A subgroup relay test shall be performed which shall include the energization/de-energization of each subgroup relay and verification of the OPERABILITY of each subgroup relay.
Relays exempt from testing ag during plant operation shall be limited to only those relays associated with plant equipment which g-cannot be operated during plant operation.
Relays not testable during plant operation shall be tested during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless tested during the previous 6 months.
t 5
(5) Actuated equipment'only; does not result in CIAS.
~
8 (6) At least once per Refueling Interval.
With irradiated fuel in the storage pool.
v m
l*
JN$TRVWTNTAT!ON 3/4. 3. 3 MONITORING !NSTRJHENTATION RA01 ATION MDNITORIN3 !NSTRUWENTATION L1MITIN3 CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in specified limits." Table 3.3 6 shall be OPERABLE with their alarm / trip setpoin A' PLICA!!LITY: As shavn in Table 3.3 8.
AEIl03:
With a radiation tonitoring channel alart/ trip a.
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,value 4
b.
ACTION shown jn Table 3.3 8.With one or more radiation monitor The provisions of Specifications 3.0.3 and 3.0.4 are not app 1feable.
c.
i
$URVE1LLAN:E REOUIREMENTS 4.3.3.1 Each radiation monitoring instrueentation channel shall be cener.strated OPERABLE by the perforeance of the CHANNEL CHECK. CHANNEL l
CALIBRATION and CHANNEt FUNCTIONAL TEST operations for the MDDEs a frequencies shown in Table 4.3 3.
" Continuous conttoring and sampling of the containment purge exhaust d frem the pur and 42 inch)ge stack shall be provided for the low and high volume (8 inch Containment airborne monitor 3RT-78041 of 3RT*7807 outa$ing redia shall perfore these functions prior to initial c ae.
sanp From initial criticality 16 the starty following the first refueling outspe r
cality.
centainment airborne coniter 3RT-7804 3 and associated sL:pling sedia shal perform the above required functions.
{
4 0
\\_
" ~ " ~ ~ ~
~
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the reactor protective and Engineered Safety Features Actuation System instrumentation and bypasses ensure that 1) the associated Engineered Safety Features Actuation System e. tion and/or reactor 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 is maintained to permit a channel to be out of service for testing or maintenance, and 4 capability is available from diverse param)eters. sufficient system functional 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 integratcd operation of each of these systems is consistent with the assumptions used in the accident analyses.
When a protection channel of a given process variable becomes inoperable, the inoperable channel may be placed in bypass until the next Onsite Review Committee meeting at which time the Onsite Review Committee will review and document their judgment concerning prolonged operation in bypass, channel triplcea/orrepair.
The goal shall be to return the inoperable channel to and serv s soon as practicable but in no case later than during the next COLD SHUTDOWN.
This approach to bypass / trip in four channel protection systems is consistant with the applicable criteria of IEEE Standards 279, 323, 344 and 384.
The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications of power level, RCS flow rate, axial flux shape, radial peaking factors and CEA deviation penalties.
Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inadvertent misloading of addressable constants into the CPCs is unlikely.
The redundancy and design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEAC's becomes in-operable.
If one CEAC is in test or inoperable, verification of CEAC position is performed at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If the second CEAC fails, the CPC's will use DNBR and LPD penalty factors, which restrict reactor operation to some maximum fraction of RATED THERMAL POWER.
If this maximum fraction is exceeded e reactor trip will occur.
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.
The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
The quarterly frequency for the CHANNEL FUNCTIONAL TESTS for these systems is based on the analyses presented in the NRC approved topical report, CEN-327, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented.
The measurement of response time at the specified frequencies provides assurance that the reactor protective and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.
SAN ONOFRE - UNIT 3 8 3/4 3-1 AMENDMENT NO.90 1
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUAT10N SYSTEM INSTRURENTATION (Continued)
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.
Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement sensors with certified response times.
l D
SAN ONOFRE - UNIT 3 8 3/4 3-la AMENDMENT NO. 36 l
/g N49jo bNITED 3T ATE s
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[,
'3 g
NOCLE AR HEGULATORY COMMISSION 5
. E WA$6 ilNG TON, D. C. 20%5
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.....s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENEMENT NO.101TO FACILITY OPERATING LICENSE NO. NPF-10 AND AMENbMENT NO.90 TO FACILITY OPERATING LICENSE NO. NPF-15 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHE1M, CALIFORNIA SAN ONOFRE NUCLEAR GENERATING STAT 10N, UNITS 2 AND 3 DOCKET NOS. 50-361 AND 50-362
1.0 INTRODUCTION
By letter dated August 30, 1991, Southern California Edison Company, et al.
(SCE or the licensee) submitted a request for changes to the Technical Specifications (TS) for San Onofre Nuclear Generating Station, Unit Nos. 2 and 3.
The proposed changes would revise TS 3/4.3.1, " Reactor Protection System Instrumentation," and TS 3/4.3.2, Engineered Safety Features Actuation System Instrumentation." These amendments modify the channel functional and logic units surveillance test intervals from monthly to quarterly.
2.0 EVALUATION The proposed amendment is based on topical reports CEN-327-A and CEN-327-A Supplement 1.
Both reports were prepared by Combustion Engineering for the Combustion Engineering Owners Group (CEOG). The purpose of these reports was to evaluate the safety impact and provide justification for extending the current 30 day surveillance test interval for both RPS and ESFAS instrumenta-tion. Both reports used probability risk analysis techniques to demonstrate that the proposed surveillance interval extensions do not result in increased plant risk when compared with current technical specification requirements.
The NRC evaluation and acceptance of the topical reports is documented by a safety evaluation resort (SER) that was sent to the chairman of the CEOG on November 6, 1989.
T1e NRC found that the referenced topical reports provide an acceptable generic basis to support plant specific TS changes for extending both RPS and ESFAS channel functional test intervals from monthly to quarterly.
9204230043 920228 PDR ADOCK 05000361 P
ppg
2 The CF analysis estimattd a slight increase in RPS unavailability as a result of extending the surveillance test interval. The analysis also estimated a reduced core melt frequency based on a reduction in surveillance test induced transients. The overall effect of the proposed change on safety was determined to be negligible. The result of reduced ESFAS testing on core melt frequency was found to be similar to that for RPS.
The San Onofre Units 2 and 3 Technical Specifications in section 3/4.3.1 provide instrumentation operability and surveillance requirements for the RPS.
Technical Specification 4.3.1.1 and Table 4.3-1 specify the modes and frequency for the performance of channel -hecks, channel functional tests and channel calibration for each RPS channel.
The staff SER for CEN-327 concluded that the CE report did not address the effects of drift in both the sensors or instrument strings. The effects of drift are plant specific and therefore should be included with each iw sidual plant analysis. As stated in the generic SER, each licensee should confirm that they have reviewed drift information including as found and as lef t values for each instrument channel involved and have determined that the drift occurring in that channel will remain bounded by the setpoint methodology for the extended surveillance interval.
Additionally, the licensee should maintain records of the setpoint calculations and associated data to support future staff audits.
The licensee stated that the calibration of transmitters and signal processing equipment is normally done at each refueling interval and is not affected by the proposed increase in the functional test surveillance interval. The surveillance test calibration interval for this equipment is not being changed.
However, the licensee stated that an increase from monthly to quarterly for the channel functional test does affect the bistable trip units. The licensee performed an analysis of the bistable trip unit drift records including as found and as left values. The licensee concluded that the trip setpoints will be within the established pass / fail cr, sria when testing is performed quarterly.
The staff requested the licensee to confirm that for any pro)osed extension of monthly functional test intervals, the bases for the 24 monti calibration surveillance interval will not be compromised. The licensee stated that the surveillance and corrective maintenance history indicates that problems are identified as a result of the shif t channel checks and during routine monitoring of plant parameters. Since the monthly functional test involves the injection of simulated signals into the RPS/ESFAS logic, any failure relating to instru-ment calibration would not be detected by this testing methodology. However, a channel check may reveal information identifying a calibration related problem.
The channel check surveillance is not being revised by the licensee and will continue to be performed once per shift.
i
- The CEOG topical report addressed the s <nnel funtional test frequency for all the functional units referenced in Tabh
.3 1 except for the manual reactor trip, reactor trip breakers, and seismic high trip. The manual reactor trip functional test is currently specified to be performed every refueling outage and is not being resised.
The reactor trip breakers channel functional test interval will remain 18 months. Functional units for the reactor protection system logic, core protection calculators, and control element assembly calculators were not identified as specific dominant cut sets in
.EN 327-A but were considered in the analysis. Although seismic high is not iisted as a cut set for RPS, the licensee has proposed that the seismic high functional test frequency be revised from monthly to quarterly based on the similarity of bistable design with the loss of load trip function. The licensee stated that the analysis used to justify the loss of load trip function functional test is also applicable to the seismic high functional unit.
This conclusion was confirmed by CE.
Table 4.3-2 specifies the functional test surveillance requirements for the ESFAS. Topical report CEN 327-A addressed all the functional units referenced in Table 4.3-2 except for the containment cooling actuation signal (CCAS), the control room isolation signal (CRIS), the toxic gas isolation signal (TGIS),
the fuel handling isolation signal (FHIS), and the containment purge isolation signal (CPIS). The licensee proposed to extend the surveillance interval for CCAS from monthly to quarterly with CRIS, TGIS, FHIS, and CPIS remaining as specified in the current TS, Again, C;rt-327-A does not specifically address CCAS but the licensee stated that the CCAS and the SIAS share the same type of bistable and are designed similarly. Therefore, the licensee stated that the associated analysis justifying a functional test interval extension for SIAS is also applicable to CCAS. This conclusion was confirmed by CE.
3.0 CONCLUSION
The RPS/ESFAS test interni esalvation presented in CEN-327-A and CEN-327-A Supplement I developed a f wit tree model for the four classes of RTS and three classes of ESFAS design.
Ea 6 model addressed common mode failures, operator errors, reduced redundancy, and random component f ailures. These models were used to evaluate the RTS and ESFAS availability based on a 30 day and 90 day test interval. The CE analysis (CEN-327-A and CEN-327 Supplement 1) concludes that there would be a slight increase in RPS unavailability as a_ result of extending the test interval from monthly to quarterly.
The analysis also concluded that reducing the test interval would reduce the scram and core melt frequency based on the expected reduction in test induced transients / scrams.
The staff found these estimates to be acceptable. The staff SER for CEN-327-A found the overall impact of reduced testing intervals on safety to be negligible. The results of the CE enalysis regarding reduced ESFAS testing on core melt frequency was found to similar to RPS.
The staff SER for CEN-327-A required the licensee to evaluate the effects of drift on the proposed functional test interval extension. The licensee
. ~.. _ _. _. _ _, _ _ _ _ _ _ _
m...
-4 reviewed the drift data (as left, as found) for the affected instrumentation and determined that the projected drift is bounded by the current setpoint calculations.
The evaluation results are acceptable to the staff.
The functional units not specifically referenced in CEN-327-A for surveillance interval extension but proposed by the licensee to be included in the TS arnendment (seismic high and containment cooling actuation signal) utilize similar bistables and/or design when corrpared to functional units analyzed by CEN-327-A. The staff finds the basis for including the additional functional units acceptable.
Based on the above, the staff finds the licensee proposal to incorporate the quarterly surveillance test intervals for RPS and ESFAS instrumentation as referenced by CEN-327-A and CEN-32M Supplement 1 to be acceptable.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATICN The atendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.
The NRC staff has determined that the amendments involve no sigt.ificant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupaticnal radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (56 FR 49926). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impect statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuante of the amendrnent will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
Clifford K. Doutt Lawrence E. Kokajko Date:
February 28, 1992