ML20195C732
| ML20195C732 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/16/1986 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20195C730 | List: |
| References | |
| TAC-59885, TAC-59886, TAC-59887, TAC-59888, TAC-59889, TAC-59890, TAC-59891, TAC-59892, TAC-59893, TAC-59894, NUDOCS 8605300495 | |
| Download: ML20195C732 (88) | |
Text
7 f >SKth,Iog UNITED STATES y'
g NUCLEAR REGULATORY COMMISSION ns j
WASHINGTON, D. C. 20655 s
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SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF-ANAHEIM,' CALIFORNIA DOCKET NO. 50-361 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 47 License No. NPF-10 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application 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 October 9, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 Commission; C.
There is reasonable assurance:
(1)thattheactivitiesauthorized 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; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; 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.
.s.
4
. 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. 47, at~e hereby incorporsted in the license.
SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective immediately and is to be fully implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION eorge
Attachment:
Changes to the Technical Specifications Date of Issuance: May 16, 1986 1
g l
May 16, 1986
. ATTACHMENT TO LICENSE AMENDMENT NO. 47 FACILITY OPERATING LICENSE NO. NPF-10 DOCKET NO. 50-361 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contains vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.
Amendment Page Overleaf Page 2-2 2-1 2-3 2-4 2-5 2-6 2-7 B 2-2 B 2-1 B 2-8 B 2-7 3/4 1-4 3/4 1-3 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-7a 3/4 2-8 3/4 2-8a 3/4 2-8b 3/4 2-8c 3/4 3-6 3/4 3-5 3/4 3-7 3/4 3-7a 3/4 3-9 3/4 3-8 3/4 3-9a 3/4 3-9b B 3/4 2-3 B 3/4 2-4 B 3/4 3-1 B 3/4 3-1A B 3/4 3-2 6-7 6-8 6-14 6-13 6-14a
_c 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS s
i 2.1 SAFETY LIMITS 2.1.1 REACTOR CCRE ON8R 2.1.1.1 The DNBR of the reactor core shall be maintained greater than or equal to 1.31.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the DN8R of the reactor has decreased to less than 1.31, be in HOT STAN08Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
PEAK LINEAR HEAT RATE 2.1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained less than or equal to 21.0 kw/ft.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21.0 kw/ft, be in HOT STAN08Y within I hour, and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
APPLICA8ILITY: MODES 1, 2, 3, 4 and 5.
ACTION:
MODES 1 and 2 Whenever the Reactor Coolant Systra pressure has exceeded 275O psia, be in HOT STAN08Y with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
MODES 3, 4 and 5 i
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.
SAN ONOFRE-UNIT 2 2-1 AMENOMENT NO. 32 l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM' SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specifica-tion 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
L SAN ONOFRE-UNIT 2 2-2 AMEN 0 MENT NO. 47
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TABLE 2.2-1 s
]
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 5
k FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES g
1.
Manual Reactor Trip Not Applicable Not Applicable 2.
Linear Power Level - High -
Four Reactor Coolant Pumps 5 110.0% of RATED THERMAL POWER
$ 111.3% of RATED THERMAL POWER Operating 3.
Logarithmic Power Level - High (1) 5 0.89% of RATED THERMAL POWER 5 0.96% of RATED THERMAL POWER 4.
Pressurizer Pressure - High 5 2382 psia 5 2389 psia 5.
Pressurizer Pressure - Low (2) 1 1806 psia 1 1763 psia
]
6.
Containment Pressure - High 5 2.95 psig 5 3.14 psig 7.
Steam Generator Pressure - Low (3) 1 729 psia 1 711 psia 8.
Steam Gererator Level - Low 1 25% (4) 1 24.23% (4) 9.
Local Power Density - High (5) 5 21.0 kw/ft 5 21.0 kw/ft 10.
DNBR - Low
> 1.31 (5) 1 1.31 (5) 11.
Reactor Coolant Flow - Low y
a) DN Rate 5 0.22 psid/sec (6)(8) 5 0.231 psid/sec (6)(8) g b) Floor 1 13.2 psid (6)(8) 1 12.1 psid (6)(8) y c) Step 5 6.82 psid (6)(8)
$ 7.231 psid (6)(8)
E 12.
Steam Generator Level - High 5 90% (4) 5 90.74% (4) 5 13.
Seismic - High 5 0.48/0.60 (7) 5 0.48/0.60 (7) 14.
Loss of Load Turbine stop valve closed Turbine stop valve closed
i TABLE 2.2-1 (Continued)
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jg REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS o
TA8tE NOTATION.
k (1) Trip may be manually bypassed above 10 *% of RATED THERMAL POWER; bypass shall be automatically q'
removed when THERMAL POWER is less than or equal to 10 4% of RATED THERMAL POWER.
SE (2) Value may be decreased manually, to a minimum value of 300 psia, as pressurizer pressure is reduced, -
- q provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is m
increased until the trip setpoint is reached.
Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(4) % of the distance between steam generator. upper and low level instrument nozzles.
(5) As stored within the Core Protection Calculator (CPC).
Calculation of the trip setpoint includes gi measurement, calculational and processor uncertainties, and dynamic allowances.
Trip may be manually bypassed below 10 *% of RATED THERMAL POWER; bypass shall be automatically removed _when a
THERMAL POWER is greater than or equal to 10 4% of RATED THERMAL POWER.
The approyed DNBR limit accounting for use of HID-2 grids is 1.31.
(6) DN RATE is the maximum decrease rate of the trip setpoint.
FLOOR is the minimum value of the trip setpoint.
STEP is the amount by which the trip setpoint is below the input signal unless limited by DN Rate or floor.
(7) Acceleration, horizontal / vertical, g.
(8) Setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.
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f 2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforetion which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting feel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) main-taining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kw/ft which will not cause fuel centerline melting in any fuel rod.
First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed " departure from n'cleate boiling" (DN8). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DN8 and'the location of DN8 for axially unifore and 1
non-uniform heat flux distributions.
The local DN8 ratio (DNBR), defined as the ratio of the predicted DN8 heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DN8. The minimum value of DN8R during normal operational occurrences is limited to 1.31 l
for the CE-1 correlation and is established as a Safety Limit.
Second, operation with a peak linear heat rate below that which would cause fuel centerline melting raintains fuel rod and cladding integrity.
Above this peak linear heat race level (i.e., with some melting in the center), fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume changes which accompany the solid to liquid phase change are signif'icant and require accommodation. Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
Limiting safety system settings for the Low DN8R, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DN8R and kw/ft margin are specified such that there is a high degree c,f confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.
i l
l SAN ON0FRE-UNIT 2 8 2-1 AMENDMENT No. 32 i
j SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III, 1971 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and asso-ciated code requirements.
The entire Reactor Coolant System was hydrotested at 3125 psia to demon-strate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
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 DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.31 and 21.0 kw/ft, respectively.
Since these trips u.e digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment.
The Seismic-High trip is generated by an open contact signal from a force balance contact device which is likewise not subject to analog type drifts.
The Allowable Values for these trips are therefore the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density -
High trips include the measurement, calculational and" processor uncertainties and dynamic allowances as defined in CEN-147(S)-P, " Functional Design Specification for a Core Protection Calculator," January, 1981; CEN-148(S)-P,
" Functional Design Specification for a Control Element Assembly Calculator,"
January,1981; CEN-149(S)-P "CPC/CEAC Data Base Document", January,1981, and CEN-175(S)-P " SONGS 2 Cycle 1 CPC and CEAC Data Base Document", August, 1981.
Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
e
.J SAN ONOFRE-UNIT 2 B 2-2 AMENDMENT NO. 47
i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR-Low (Continued)
This illustrates the methodology used for conversion of any DNBR penalty into a format that is useable and addr'ssable in both CPC and COLSS.
e The addressable constants BERR1 and EPOL2 are also used to accommodate the DN8R rod bow penalties listed in Technical Specification 4.2.4.4.
Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against a reactor coolant pump sheared shaft event and a two pump opposite loop flow coastdown event.
side of either steam generator goes below a variable setpoint.A trip is This variable setpoint stays a set amount below the pressure differential unless limited by a set maximum decrease rate or a set minimum value.
The specified setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNBR safety limits under the stated conditions.
Seismic - High The Seismic - High trip is provided to trip the reactor in the event of an earthquake which exceeds 60% of the Safe Shutdown Earthquake level.
trip's setpoint does not correspond to a safety limit and no credit was taken This in the accident analyses for operation of this trip.
Loss of load The Loss of Load trip is provided to trip the reactor when the turbine is tripped above a predetermined power level.
This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip setting enhances the overall reliability of the Reactor Protection System.
Steam Generator Level-High The Steam Generator Level-High trip is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over.
trip's setpoint does not correspond to a Safety Limit and no credit was taken This in the accident analyses for operation of this trip.
at the specified trip setting enhances the overall reliability of the ReactorIts functional c Protection System.
SAN ONOFRE-UNIT 2 8 2-7 AMEN 0 MENT NO. 21
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DELETED INTENTIONALLY 1
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47 SAN ONOFRE-UNIT 2
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e REACTIVITY CONTROL SYSTEMS SHUT 00W MARGIN - T,,
LESS THAN 6R EQUAL TO 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUT 00W MARGIN shall be greater than or equal to 3.0% delta k/k.
l APPLICA81LITY: MODE 5.
ACTION:
With the SHUT 00W MARGIN less than 3.0% delta k/k, immediately initiate and l
continue boration at greater than or equal to 40 gpa of a solution containing greater than or equal to 1720 ppe boron or equivalent until the required SHUT 00W MARGIN is restored.
SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 3.0% delta k/k:
a.
Within one hour after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.
If the inoperable CEA is immovable or untrippable, the above required SHUTOOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).
4 b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration cf the following factors:
1.
Reactor coolant system boron concentration, 2.
CEA position, 3.
Reactor coolant system average temperature, i
4.
Fuel burnup based on gross thermal energy generation, J
5.
Xenon concentration,and j
6.
Samarium concentration.
7.
Whenever the reactor coolant level is below the hot leg i
centerline, one and only one charging pump shall be operable; by verifying that power is removed from the remaining charging pumps.
l SAN ON0FRE-UNIT 2 3/4 1-3 AMENOMENT NO. 28 1
j i
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION F0,R OPERATION t
3.1.1.3 The moderator temperature coefficient (MTC) shall be:
~4 a.
tess positive than 0.5 x 10 delta k/k/*F whenever THERMAL POWER is 1 70% of RATED THERMAL POWER, or Less positive than 0.0 delta k/k/*F whenever THERMAL POWER is > 70%
of RATED THERMAL POWER, and
-4 b.
Less negative than -3.3 x 10 delta k/k/*F at RATED THERMAL POWER.
APPLICABILITY:
MODES 1 and 2*#
ACTION:
With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.
I 4.1.1.3.2 The MTC shall be determined at the following frequencies and i
THERMAL POWER conditions during each fuel cycle:
a.
Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
b.
At any THERMAL POWER, within 7 EFPD of reaching 40 EFP0 core burnup.
c.
At any THERMAL POWER, within 7 EFPD of reaching 2/3 of_ expected core burnup.
1 "With K,ff greater than or equal to 1.0.
- See Special Test Exception 3.10.2.
l SAN ONOFRE-UNIT 2 3/4 1-4 AMEN 0 MENT N0. 47
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t POWER DISTRIBUTION LIMITS
)
3/4.2.4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DN8R margin shall be maintained by one of following methods:
a.
Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR (when COLSS is in service, and either one or both CEACs are operable); or b.
Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR decreased by 13.0% RATED THERMAL POWER (when COLSS is in service and neither CEAC is operable): or c.
Operating within the region of acceptable operation of Figure 3.2-1 using any operable CPC channel (when COLSS is out of service and either one or both CEACs are operable); or d.
Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC channel (when COLSS is out of service and neither CEACs is operable).
APPLICABILITY:
MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the ONBR margin not being maintained, as indicated by:
(1) COLSS calculated core power exceeding the appropriate COLSS calculated operating limit, or (2) With COLSS out of service, operation outside the region of acceptable operation of Figure 3.2-1 or 3.2-2.
Within 15 minutes initiate corrective action to restore the DNBR to within its limits, and either:
a.
Restore the DNBR to within its limits within one hour, or b.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on any OPERABLE ONBR channel, is within the limit shown on Figures 3.2-1 or 3.2-2, as applicable.
4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR.
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-0.1 0.0 0.1 0.2 0.3 AX1AL SHAPE INDEX Figure 3.2-1 DNBR OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS
- COLSS OUT OF SERVICE
- ONE OR BOTH CEACS OPERABLE SAN ON0FRE-UNIT 2 3/4 2-7 AMENDMENT NO. 47 l
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- COLSS OUT OF SERVICE
- BOTH CEACS IN0PERABLE SAN ONOFRE-UNIT 2 3/4 2-8 AMENDMENT NO. 47
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TABLE 3.3-1 (Continued) i ACTION STATEMENTS With a chaanel process meas'urement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:
Process Measurement Circuit functional Unit Bypassed 1.
Linear Power Linear Power Level - High (Subchannel or Linear)
Local Power Density - High DNBR - Low 2.
Pressurizer Pressure - High Pressurizer Pressure - High Local Power Density - High DN8R - Low 3.
Containment Pressure - High Containment Pressure - High (RPS)
Containment Pressure - High (ESF) 4 Steam Generator Pressure -
Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5.
Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator aP (EFAS) 6.
Core Protection Calculator Local Power Density - High DNBR - Low ACTION 3 -
With the number of channels OPERA 8LE one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:
a.
Verify that one of the inoperable channels has been bypassed and place the other channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and b.
All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below:
Process Measurement Circuit Functional Unit Bypassed / Tripped 1.
Linear Power Linear Power Level - High (Subchannel or Linear)
Local Power Density - High DNBR - Low l
(
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l TABLE 3.3-1 (Continued)
ACTION STATEMENTS 2.
Pressurizer Pressure -
Pressurizer Pressure - High High Local Power Density - High DNBR - Low 3.
Containment Pressure -
Containment Pressure - High (RPS)
High Containment Pressure - High (ESF) 4.
Steam Generator Steam Generator Pressure - Low Pressure - Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5.
Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6.
Core Protection Local Power Density - High Calculator DNBR - Low STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST.
Subsequent STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied.
ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 6 -
a.
With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, each CEA is verified to be within 7 inches (indicated position) of all other CEA's in its group.
After 7 days, operation may continue provided that Action 6.b is met.*
b.
With both CEACs inoperable, operation may continue provided that:*
1.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the DNBR margin required by Specification 3.2.4.b (COLSS in service) or Specification 3.2.4.d (COLSS out of service) is satisfied.
- Note:
Requirements for CEA position indication given in Technical l
Specification 3.1.3.2.
l SAN ON0FRE-UNIT 2 3/4 3-6 AMENDMENT N0. 47
TABLE 3.3-1 (Continued)
TABLE NOTATION 2.
Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a)
All full length and part length CEA groups are withdrawn to and subsequently maintained at the
" Full Out" position, except during surveillance testing pursuant to the requirements of Specifica-tion 4.1.3.1.2 or for control when CEA group 6 may be inserted no further than 127.5 inches withdrawn.
b)
The "RSPT/CEAC Inoperable" addressable constant t in the CPC's is set to indicate that both CEAC's are inoperable.
c)
The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the "Off" mode except during CEA group 6 motion permitted by a) above, when the CEDMCS may be operated in either the " Manual Group" or " Manual Individual" mode.
3.
At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, all full length and part length CEA's are verified fully withdrawn except during surveillance testing pursuant to Specification 4.1.3.1.2 or during insertion of CEA group 6 as per-mitted by 2.a) above, then verify at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> that the inserted CEA's are aligned within 7 inches (indicated position) of all other CEA's in its group.
I ACTION 7
- With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 7A
- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.
i SAN ON0FRE-UNIT 2 3/4 3-7 AMENDMENT N0. 47
.~,
DELETED INTENTIONALLY l
l 1
SAN ONFRE-UNIT 2 3/4 3-7a AMEN 0 MENT NO. 47
TABLE 3.3-2 NE 1
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE ilHES E
S$
FUNCTIONAL UNIT RESPONSE TIME c:
jc 1.
Manual Reactor Trip tiot Applicable 2.
Linear Power Level - liigh 1.0.40 seconds" 3.
Logarithmic Power Level - liigh 3 0.45 seconds" 4.
Pressurizer Pressure - liigh i 0.90 seconds 5.
Pressurizer Pressure - Low 1 0.90 seconds 6.
Containment Pressure - lligh 5 0.90 seconds 7.
Steam Generator Pressure - tow
$ 0.90 seconds a
N' 8.
Steam Generator Level - tow
$ 0.*J0 seconds m
9.
Local Power Density - High a.
Neutron Flux Power from Excore Neutron Detectors
< 0.68 seconds
- b.
CEA Positions i 0.6n seconds ^
a c.
CEA Positions: CEAC Penalty factor 30.53 seconds
- 10. DNBR - Low i
il a.
Neutron Flux Power from Excore Neutron Detectors
< 0.68 seconds" b.
CEA Positions E 0.68 seconds **
c.
Cold Leg Temperature i 0.60 seconds ##
I h{
d.
Ilot Leg Temperature
_ 0.68 seconds ##
)
25 e.
Primary Coolant Pump Shaft Speed
$ 0.6ft seconds #
gi f.
Reactor Coolant Pressure from Pressurizer 5 0.611 seconds 3;
g.
CEA positions:
CEAC Penalty factor 1 0.S3 seconds ei l
TABLE 3.3-2 (Continued)
'E REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES E%
FUNCTIONAL UNIT RESPONSE TIME Aa 11.
Steam Generator Level - High Not Applicable 5
12.
Reactor Protection System Logic Not Applicable N
13.
Reactor Trip Breakers Not Applicable 14.
Core Protection Calculators Not Applicable 15.
CEA Calculators Not Applicable 16.
Reactor Coolant Flow-Low 0.9 sec 17.
Seismic-High Not Applicable mk 18.
Loss of Load Not Applicable w
S
- Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
- Response time shall be measured from the onset of a single CEA drop.
- Response time shall be measured using a simulated Reactor Coolant Pump coastdown.
5
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1 DELETED INTENTIONALLY l
1 i
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1 1
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1 4
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SAN ONOFRE-UNIT 2 3/4 3-9a AMEN 0 MENT NO. 47 1
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SAN ONOFRE-UNIT 2 3/4 3-9b AMENOMENT NO. 47 1
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POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TILT - Tq (Continued) 4
~
T is the peak fractional tilt amplitude at the core periphery q
g is the radial normalizing factor i
e is the azimuthal core location O is the azimuthal core location of maximum tilt g
P
/P is the ratio of the power at a core location in the presence tilt untilt of a tilt to the power at that location with no tilt.
3/4.2.4 DNBR MARGIN l
The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety.
analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences, of which the loss of flow transient is the most limiting.
Opera-tion of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core i
Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits. The COLSS performs this function by continuously monitoring the core 4
power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR.
The COLSS calculation of core power operating limit based on the minimum DNBR limit includes appropriate penalty factors which provide, with a 95/95 probability / confidence level, that the core power limit calculated by COLSS (based on the minusum DNBR limit) is conservative with respect to the actual core power limit.
These penalty factors are determined
]
from the uncertainties associated with planar radial peaking measurement, engineering design factors, state parameter measurement, software algorithm modelling, computer processing, rod bow and core power measurement.
Parameters required to maintain the margin to DNB and total core power are also monitored by the CPCs.
In the event that the COLSS is not being used, the DNBR margin can be maintained by monitoring with any operable CPC channel so that the DNBR remains above the predetermined limit as a function of Axial Shape Index.
The above listed uncertainty penalty factors are also included in the CPCs, which assume a minimum of 20% of RATED THERMAL POWER.
The 20% RATED THERMAL POWER threshold is due to the excore neutron flux detector system being less accurate below 20% core power.
Core noise level at low power is too large to obtain usable detector readings.
The additional uncertainty terms taken into account in the CPCs for transient protection are removed from Figures 3.2-1 and 3.2-2 since the curves are intended to monitor the LC0 only during steady state operation.
SAN ON0FRE - UNIT 2 8 3/4 2-3 AMENDMENT NO. 47
POWER DISTRIBUTION LIMITS BASES ONBR Margin (continued)
A DNBR penalty factor has been included in the COLSS and CPC DNBR calculation to accommodate the effects of rod bow.
The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly.
Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.
Conversely, lower burnup assemblies will experience less rod bow.
In design calculations, the penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak.
A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses.
3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses.
2.4.2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.
4 3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of values used in the safety analyses.
l SAN ON0FRE-UNIT 2 B 3/4 2-4 AMENDMENT NO. 47
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 action 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) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the 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 trip, and/or 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 applicable criteria of IEEE Standards 279, 323, 344 and 384.
4 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 i
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 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 frequencies are sufficient to demonstrate this capability.
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 ON0FRE - UNIT 2 8 3/4 3-1 AMENDMENT N0. 47
5 e
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATI0li (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.
SAN ON0FRE - UNIT 2 B 3/4 3-la AMENDMENT N0, 47 I
I l
INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING ALARM INSTRUMENTATION The OPERABILITY of the radiation monitoring alarm channels ensures that
- 1) the radiation levels are continually measured in the areas served by the individual channels; 2) the alarm or automatic action 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 the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November, 1980.
3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum comple-ment of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.
3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event 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 facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100.
The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earth-quakes," April 1974.
3/4.3.3.4.
METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is 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 and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite l
Meteorological Programs," February 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remnte shutdown instrumentation in Panel LO42 ensures that sufficient capability is available to permit shutdown and main-tenance of HOT STANDBY of the facility from locations outside of the control SAN ON0FRE-UNIT 2 8 3/4 3-2
s
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ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.1.4 The OSRC shall meet at least once per calendar month and as convened by the OSRC Chairman or his designated alternate.
QUORUM 6.5.1.5 The minimum quorum of the OSRC necessary for the performance of the OSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.
RESPONSIBILITIES 6.5.1.6 The Onsite Review Committee shall be responsible for:
a.
Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Nuclear Safety Group (NSG).
b.
Review of events requiring 24-hour written notification to the Commission.
Review of unit operations to detect potential nuclear safety hazards.
c.
d.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Station Manager or the NSG.
Res ew and documentation of judgment concerning prolonged operation in e.
bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last OSRC meeting.
SAN ON0FRE-UNIT 2 6-7 AMEN 0 MENT NO. 47
v
^
ADMINISTRATIVE CONTROLS AUTHORITY 6.5.1.7 The Onsite Review Committee (OSRC) shall Render determinations in writing with regard to whether or not items a.
considered under 6.5.1.6(a) above constitute unreviewed safety questions.
b.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Manager of Nuclear Operations and NSG of disagreement between the OSRC and the. Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The Onsite Review Committee shall maintain written minutes of each OSRC meeting that, at a minimum, document the results of all OSRC activities performed under the responsibility and authority provisions of these technical specifications.
Copies shall be provided to the Nuclear Safety Group.
6.5.2 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.2.1 The Station Manager shall assure that each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, is prepared by a qualified individual / organization.
Each such procedure, and ~ changes thereto, shall be reviewed by an individual / group other than the individual / group which prepared the procedure, or changes thereto, but who may be from the same organization as the individual / group which prepared the procedure, or changes thereto.'
6.5.2.2 Proposed changes to the Appendix "A" Technical Specifications shall be prepared by a qualified individual / organization.
The preparation of each proposed Technical Specifications change shall be reviewed by an individual /
group other than the individual / group which prepared the proposed change, but who may be from the same organization as.the individual / group which prepared the proposed change.
Proposed changes to the Technical Specifications shall be approved by the Station Manager.
6.5.2.3 Proposed modifications to unit nuclear safety related structures, systems and components shall be designed by a qualified individual /
organization.
Each such modification shall be reviewed by an individual / group other than the individual / group wnich designed the modification, but who may be from the same organization as the individual / group which designed the modifi-cation.
Proposed modifications to nuclear safety related structures, systems and components shall be approved prior to implementation by the Station Manager; or by the Manager, Technical as previously designated by the Station Manager.
l l
SAN ONOFRE-UNIT 2 6-8 Amendment No. 4 i
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e ADMINISTRATIVE CONTROLS 7
i 6.6 REDORTABLE OCCURRENCE ACTION
- 6. 6.1 The following actions shall be taken for REPORTA8LE OCCURRENCES:
a.
The Commission shall be notified and/or a report submitted pursuant to the requirements of 5;ecification 6.9.
b.
Each 2:90RTASLE OCCURRENCE recuiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Ccm assion shall be caviewed by the OSRC and submitted to the NSG ard the Manager of Nuclear Operations.
6.7 SAFETf LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a.
The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.
The Manager of Neclear Operations and the NSG Chairman shall be notified witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the OSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation Jpon facility compon*ents, systems or structures, and (3) corrective action taken to prevent recurre'nce.
c.
The Safety Limit Violation Report shall be submitted to the Commission, the Manager of Nuclear Operations and the NSG within 14 days of the violation.
d.
Critical operation of the unit shall not be resumed until authorized by the Commission.
6.8 FROCEDURES At40 PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
The applicable procedures recommended in Appendix "A" of Regulatory a.
Guide 1.33, Revision 2, February 1978.
b.
Refueling operatior.s.
c.
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
e.
Emergency Plan implementaticn.
- f., Fire Protection Program implemestation.
SAN ONOFRE-UNIT 2 6-13
'~~
G-.
f ADMINISTRATIVE CONTROLS 1
l g.
PROCESS CONTROL PROGRAM implementation.*
i h.
OFFSITE DOSE CALCULATION MANUAL implementation.
i i.
Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15 Rev. 1, February 1979.
j.
Modification of Core Protection Calculatcr (CPC) Addressable Constants.
These procedures should include provisions to assure that sufficient margin is maintained in CPC Type I addressable constants to avoid ex-cessive operator interaction with the CPCs during reactor operation.
NOTE:
Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change 1~
Procedure," CEN-39(A)-P that has been determined to be applicable to the facility.
Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.
4 6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be approved
[
by the Station Manager; or by (1) the Deputy Station Manager, (2) the Manager, i
Operations, (3) the Manager, Maintenance, (4) the Manager, Technical, or (5) the Manager, Health Physics as previously designated by the Station Manager; j
prior to implementation and shall be reviewed periodically as set forth in administrative procedures.
l 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
l a.
The intent of the original procedure is not altered, b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
4 l
c.
The change is documented, reviewed and approved by the Station Manager; or by (1) the Deputy Station Manager, (2) the Manager, Operations, (3) the Manager, Maintenance, (4) the Manager, Technical, 1
or (5) the Manager, Health Physics as previously designated by the Station Manager; within 14 days of implementation.
6.8.4 The following programs shall be established, implemented, and maintained:
l a.
Primary Coolant Sources Outside Containment
+
t A program to reduce leakage from those portions of systems outside 1
containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.
The systems include the high pressure safety injection recirculation, the shutdown cooling system, the reactor coolant sampling system
(
(post-accident sampling piping only), the containment spray system, j
the radioactive waste gas system (post-accident sampling return 1
s i
t SAN ONOFRE-UNIT 2 6-14 AMENDMENT NO. 47
_ _ ~. _ _ _. _, _ _ _ _ _
a v
s A
t ADMINISTRATIVE CONTROLS piping only) and the liquid radwaste system (post-accident sampling return piping or.ly).
The program shall include the following:
(i) Preventive maintenanc and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.
J
't i
k t
I l
+
/
t i
1 "See Specification 6.13.1 i
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SAN ONOFRE-UNIT 2 6-14a AMENDMENT NO. 47 l w-
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WASHINGTON, 0. C. 20555
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SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC CONPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET N0. 50-362 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 License No. NPF-15 I
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application 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 October 9, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 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 set forth in 10 CFR Chapter I; D.
The issuance of this license amendment will not be inimical to the.
common defense and security or to the health and safety of the public; 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.
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1 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. 36, are hereby incorporated in the license.
SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective on initial entry into the applicable MODE of Cycle 3 with the following exceptions, which are effective immediately and are to be fully inplemented within 30 days of the date of issuance:
Pages 2-2, 2-5, 2-6, 2-7, B 2-8, 3/4 2-6, 6-9, 6-15, 6-15a, B 3/4 3-1, and B 3/4 3-1A.
FOR THE NUCLEAR REGULATORY COMMISSION
/
Geor W. Knight
, Director PWR Project Directorate No. 7 Division of PWR Licensing-B
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 16, 1986 l
f
May 16, 1986
~
3-ATTACHMENT TO LICENSE AMENDMENT NO. 36 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 indicating the area of change. Also to be replaced is the following overleaf page to the amended page.
Amendment Page Overleaf Page 2-2 2-1 2-3 2-4 2-5 2-6 2-7 B 2-2 B 2-1 B 2-8 8 2-7 3/4 1-4 3/4 1-3 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-7a 3/4 2-8 3/4 2-8a 3/4 2-8b 3/4 2-8c 3/4 3-6 3/4 3-5 3/4 3-7 1
4 3/4 3-7a 3/4 3-9 3/4 3-8 i
3/4 3-9a 3/4 3-9b B 3/4 2-3 1
B 3/4 2-4 l
8 3/4 3-1 B 3/4 3-la B 3/4 3-2 6-9 6-8 1
6-15 4
6-15a
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS I
2.1 SAFETY LIMITS em3 2.1.1 REACTOR CORE DNBR 2.1.1.1 The DNBR of the reactor core shall be maintained greater than or equal to 1.31.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the DNBR of the reactor has decreased to less than 1.31, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
PEAK LINEAR HEAT RA75 2.1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained less than or equal to 21.0 kw/ft.
l APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21.0 kw/ft, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with l
the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
i APPLICABILITY:
MODES 1, 2, 3, 4 and 5.
ACTION.
1 MODES 1 and 2 Whenevor the Reactor Coolant System pressure has exceeded 2'750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.
SAN ON0FRE - UNIT 3 2-1 AMENDMENT NO. 21
'O SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY:
As shown for each channel in Table 3.3-1.
ACTION:
With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specifica-tion 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
l l
SAN ONOFRE - UNIT 3 2-2 AMENDMENT NO. 36
~
TABLE 2.2-1 z
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS S
o j
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABL'E VALUES
[
1.
Manual Reactor Trip Not Applicable Not Applicable z
Z 2.
Linear Power Level - High -
ca Four Reactor Coolant Pumps 1 110.0% of RATED THERMAL POWER 5 111.3% of RATED THERMAL POWER Operating 3.
Logarithmic Power Level - High (1) 1 0.89% of RATED THERMAL POWER 5 0.96% of RATED THERMAL POWER 4.
Pressurizer Pressure - High 1 2382 psia 1 2389 psia 5.
Pressurizer Pressure - Low (2) 1 1806 psia 1 1763 psia
}
6.
Containment Pressure - High
$ 2.95 psig 1 3.14 psig 7.
Steam Generator Pressure - Low (3) 1 729 psia 1 711 psia 8.
Steam Generator Level - Low 1 25% (4)
> 24.23% (4) 9.
Local Power Density - High (5)
$ 21.0 kw/ft i 21.0 kw/ft 10.
DNBR - Low 1 1.31 (5)
> 1.31 (5) 11.
Reactor Coolant Flow - Low a) DN Rate 1 0.22 psid/sec (6)(8) 5 0.231 psid/sec (6)(8) b) Floor 1 13.2 psid (6)(8) 5 12.1 psid (6)(8) g c) Step 5 6.82 psid (6)(8) 5 7.231 psid (6)(8) 12.
Steam Generator Level - High 5 90% (4)
$ 90.74% (4) 13.
Seismic - High 1 0.48/0.60 (7) 1 0.48/0.60 (7) 14.
Loss of Load Turbine stop valve closed Turbine stop valve closed e
TABLE 2.2-1 (Continued) h REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS g
TABLE NOTATION (1) Trip may be manually bypassed above 10-4% of RATED THERMAL POWER; bypass shall be automatically rp removed when THERMAL POWER is less than or equal to 10-4% of RATED THERMAL POWER.
E (2) Value may be decreased manually, to a minimum value of 300 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than H
or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is w
increased until the trip setpoint is reached.
Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased 4
until the trip setpoint is reached.
(4) % of the distance between steam generator upper and low level instrument nozzles.
(5) As stored within the Core Protection Calculator (CPC).
Calculation of the trip setpoint includes y
measurement, calculational and processor uncertainties, and dynamic allowances.
Trip may be manually bypassed below 10-4% of RATED THERMAL POWER; bypass shall be automatically removed when a
THERMAL POWER is greater than or equal to 10-4% of RATED THERMAL POWER.
The approved DNBR limit accounting for use of HID-2 grid is 1.31.
(6) DN RATE is the maximum decrease rate of the trip setpoint.
FLOOR is the minimum value of the trip setpoint.
STEP is the amount by which the trip setpoint is below the input signal unless limited by DN Rate or Floor.
(7) Acceleration, horizontal / vertical, g.
(8) Setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.
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SAN ONOFRE - UNIT 3 2-5 AMEN 0 MENT NO. 36 i
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SAN ONOFRE - UNIT 3 2-6 AMENOMENT NO.
36
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SAN ONOFRE - UNIT 3 2-7 AMENDMENT NO. 36
2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.,1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) main-taining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kw/ft which will not cause fuel centerline melting in any fuel rod.
First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB).
At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.
The minimum value of DNBR during normal operational occurrences is limited to 1.31 for the CE-1 correlation and is established as a Safety Limit.
Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level (i.e., with some melting in the center), fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume changes which accompany the solid to liquid phase change are significant and require accommodation.
Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.
SAN ON0FRE - UNIT 3 8 2-1 AMENDMENT NO. 21
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES
~
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III, 1971 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psta) of design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and asso-ciated code requirements.
The entire Reactor Coolant System was hydrotested at 3125 psia to demon-t strate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within i
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 DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.31 and 21.0 kw/ft, respectively.
Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment.
The Seismic-High trip is generated by an open contact signal from a force balance contact device which is likewise not subject to analog type drifts.
The Allowable Values for these trips are therefore the same as the Trip Setpoints.
l To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DN8R - Low and Local Power Density -
High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN-147(S)-P, " Functional Design Specification for a Core Protection Calculator," January,1981; CEN-148(S)-P,
{
" Functional Design Specification for a Control Element Assembly Calculator,"
i January, 1981; CEN-149(S)-P "CPC/CEAC Data Base Document", January, 1981, and j
CEN-175(S)-P " SONGS 2 Cycle 1 CPC and CEAC Data Base Document", August, 1981.
Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective l
instrumentation channels and provides manual reactor trip capability.
SAN ONOFRE - UNIT 3 8 2-2 AMENOMENT NO.36
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR-Low (Continued)
This illustrates the methodology used for conversion of any DN8R penalty into a format that is useable and addressable in both CPC and COLSS.
The addressable constants BERR1 and EPOL2 are also used to accommodate the DNBR rod bow penalties listed in Technical Specification 4.2.4.4 Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against a reactor coolant pump sheared shaft event and a two pump opposite loop flow coastdown event. A trip is initiated when the pressure differential across the primary side of either steam generator goes below a variable setpoint.
This variable setpoint stays a set amount below the pressure differential unless limited by a set maximum decrease rate or a set minimum value.
The specified setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNBR safety limits under the stated conditions.
Seismic - High The Seismic - High trip is provided to trip the reactor in the event of an earthquake which exceeds 60% of the Safe Shutdown Earthquake level.
This trip's setpoint does not correspond to a safety limit and no credit was taken in the accident analyses for operation of this trip.
Loss of Load The Loss of Load trip is provided to trip the reactor when the turbine is tripped above a predetermined power level.
This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses a
for operation of this trip.
Its functional capability at the specified trip setting enhances the overall reliability of the Reactor Protection System.
Steam Generator Level-High The Steam Generator Level-High trip is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over.
This i
trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip setting enhances the overall reliability of the Reactor j
Protection System.
1 1
SAN ON0FRE - UNIT 3 8 2-7 AMEN 0 MENT NO.11
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REACTIVITY CONTROL SYSTEMS 1
SHUT 00WN MARGIN - Tav,_ LESS THAN OR EQUAL TO 200*F i
LINITING CONDITION FOR OPERATION i
1 t
i k.1.1.2 The SHUT 00WN MARGIN shall be greater than or equal to 3.0K delta k/k. l I
APPLICA8ILITY: M00E 5.
I ACTION:
With the SHUTD0WN MARGIN less than 3.0% delta k/k, immediately initiate and l
continue boration at greater than or equal to 40 gpa of a solution containing i.
greater than or equal to 1720 ppe boron or equivalent untti the required SHUT 00WN MARGIN is restored, i
4 I
l SURVEILLANCE REQUIRENENTS i
4.1.1.2 The SHUT 00WN MARGIN shall be determined to be greater than or equal l
j to 3.05 delta k/k:
l 1
i a.
Within one hour after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.
If the inoperable CEA is immovable or untrippable, the above i
required SHUT 00WN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).
4 b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:
1 1.
Reactor coolant system boron concentration, l
2.
CEA position, j
3.
Reactor coolant system average temperature, i
4.
Fuel burnap based on gross thermal energy generation, 5.
Xenon concentration,and i
6.
Samarium concentration.
l 7.
Whenever the reactor coolant level is below the hot leg center-i line, one and only one charging pump shall be operable; by veri-j fying that power is removed from the remaining charging pumps.
k l
SAN ON0FRE-UNIT 3 3/4 1-3 AMENT > MENT NO. 17 I
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__._--__._.._,,__,___.._,m.--__.--__
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r A
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:
~4 a.
Less positive than 0.5 x 10 delta k/k/*F whenever THERMAL POWER is
$ 70% of RATED THERMAL POWER, or less positive than 0.0 delta k/k/*F l
whenever THERMAL POWER is > 70% of RATED THERMAL POWER, and
~4 b.
Less negative than -3.3 x 10 delta k/k/*F at RATED THERMAL POWER.
APPLICABILITY: MODES 1 and 2*#
ACTION:
With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its limits by confirmatory measurements.
MTC measured va. lues shall be extrapolated and/or compensated to permit direct comparison with the above limits.
4.1.1.3.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:
4 a.
Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
b.
At any THERMAL POWER, within 7 EFPD of reaching 40 EFPD core burnup.
c.
At any THERMAL POWER, within 7 EFPD of reaching 2/3 of expected core burnup.
- With K,f f greater than or equal to 1.0.
- See Special Test Exception 3.10.2.
SAN ONOFRE-UNIT 3 3/4 1-4 AMENDMENT NO. 36
J t
POWER DISTRIBUTION LIMITS 3/4.2.4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by one of following methods:
a.
Maintaining COLSS calculated core power less than or eqJa1 to COLSS calculated core power operating limit based on DNBR (when COLSS is in service, and either one or both CEACs are operable); or b.
Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR decreased by 13.0% RATED THERMAL POWER (when COLSS is in service and neither CEAC is operable): or Operating within the region of acceptable operation of Figure 3.2-1 using c.
any operable CPC channel (when COLSS is out of service and either one or both CEACs are operable); or d.
Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC channel (when COLSS is out of service and neither CEACs is operable).
APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the DNBR margin not being maintained, as indicated by:
(1) COLSS calculated core power exceeding the appropriate COLSS calculated operating limit, or (2) With COLSS out of service, operation outside the region of acceptable operation of Figure 3.2-1 or 3.2-2.
Within 15 minutes initiate corrective action to restore the DNBR to within its limits, and either:
Restore the DNBR to within its limits within one hour, or a.
b.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on any OPERABLE ONBR channel, is within the limit shown on Figures 3.2-1 or 3.2-2, as applicable.
4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR.
SAN ONOFRE-UNIT 3 3/4 2-5 AMEN 0 MENT NO.36
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i SAN ONOFRE-UNIT 3 3/4 2-6 AMENDMENT NO. 36
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-0.2 s ASI < 0.04 DNSR a 1.11 AMI + 2.31 - --
l ts-0.04 sASIs0.2 DNSR=2A0 i
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ReetON OF ACCEPTABLE OPERATION I
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-0.3
-0J
-0.1 0.0 0.1 0.2 0.3 A) GAL SHAPE IM FIGURE 3.2-1 DNBR OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS
- COLSS OUT OF SERVICE
- ONE OR BOTH CEACS OPERABLE SAN ON0FRE-UNIT 3 3/4 2-7 AMENDMENT N0. 36
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i SAN ONOFRE-UNIT 3 3/4 2-7a AMENDMENT NO. 36 9
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l REGION OF UNACCEPTABLE OPERATION i
........................:i......................u...........
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...........7........-.....p....-.....
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-0.1 0.0 0.1 0.2 0.3 AX1AL SHAPE INDEX FIGURE 3.2-2 DNBR OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR
- COLSS OUT OF SERVICE
- BOTH CEACS INOPERABLE l
SAN ONOFRE-UNIT 3 3/4 2-8 AMEN 0 MENT NO. 36
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i l.
TA8LE 3.3-1 (Continued) i ACTION STATEMENTS j
With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:
(
Process Measurement Circuit Functional Unit Bypassed l
1.
Linear Power Linear Power Level - High (Subchannel or Linear)
Local Power Density - High DNBR - Low l
l 2.
Pressurizer Pressure - High Pressurizer Pressure - High j
Local Power Density - High j
DN8R - Low 3.
Containment Pressure - High Containment Pressure - High (RPS)
Containment Pressure - High (ESF) i 4.
Steam Generator Pressure -
Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS 1 and 2)~
5.
. Steam Generator Level Steam Generator Level
' Low Steam Generator Level - High Steam Generator AP (EFAS) 6.
Core Protection Calculator Local Power Density - High DNBR - Low ACTION 3 -
With the number of channels OPERABLE one less than the Minimum r
l Channels OPERA 8LE requirement, STARTUP and/or POWER OPERATION j
may continue provided the following conditions are satisfied:
4 a.
Verify that one of the inoperable channels has been bypassed and place the other channel in the tripped 1
j condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and b.
All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below:
Process Measurement Circuit Functional Unit Bypassed / Tripped 1.
Linear Power Linear Power Level - High (Subchannel or Linear)
Local Power Density - High j
DNBR - Low i
r>
e' SAN ON0FRE-UNIT 3 3/4 3-5
f N
i j
TA8LE 3.3-1 (Continued)
ACTION STATEMENTS
{
v 2.
Pressurizer Pressure -
Pressurizer Pressure - High High Local Power Density - High DN8R - Low l
3.
Containment Pressure -
Containment Pressure - High (RPS)'
High Containment Pressure - High (ESF) 4.
. Steam Generator Pressure - Low Pressure - Low Steaa. Generator AP 1~and 2 (EFAS 1 and,2) 5.
Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6.
Core Protection Local Power Density - High Calculator DNBR - Low j
i STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST.
Subsequent j
STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERA 8LE status and the provisions of ACTION 2 are j
satisfied.
ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
1 ACTION 5 With the number of channels OPERA 8EE one less than required by the Minimum Channels OPERABLE requirement, be in'at least HOT j
STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
i ACTION 6 With one CEAC inoperable, operation may contiaue for up to a.
l 7 days provided that at least'once per 4 hoyrs, each CFA 1
i is verified to be within 7 inches (indicated position) of i
all other CEA's in its group.
After 7 days, operation may j
continue provided that ACTION 6.b is met.*
l b.
WithbothCEACsinoperable,operatio$mayconinue t
provided that:*
+
1 1.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the DNBR margin reqired by Specification 3.2.4.b (COLSS in service) or i
Specification 3.2.4.d (COLSS out of service) is satisfied.
l 5
- Note:
Requirements for CEA position. indication given in Technical 3
l Specification 3.1.3.2.
s t
{
SAN ONOFRE - UNIT 3 3/4 3-6 AMEN 0 MENT NO. 36 i
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TABLE 3.3-1 (Continued)
TABLE NOTATION 2.
Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a)
All full length and part length CEA groups are withdrawn to and subsequently maintained at the
" Full Out" position, except during surveillance testing pursuant to the requirements of Speciff-cation 4.1.3.1.2 or for control when CEA group 6 may be inserted no further than 127.5 inches withdrawn.
b)
The "RSPT/CEAC Inoperable" addressable constant in the CPC's is set to indicate that both CEAC's are inoperable.
c)
The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the "Off" mode except during CEA group 6 motion permitted by a) above, when the CEDMCS may be operated in either the " Manual Group" or " Manual Individual" mode.
3.
At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, all full length and part length CEA's are verified fully withdrawn except during surveillance testing pursuant to Specification 4.1.3.1.2 or during insertion of CEA group 6 as permitted by 2.a) above, then verify at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> that the inserted CEA's are aligned within 7 inches (indicated position) of all other CEA's in its group.
ACTION 7 With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERA 8ILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 7A -
With the number of OPERA 8LE channels one less than the Minimum Channels OPERA 8LE requirement restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.
l SAN ONOFRE - UNIT 3 3/4 3-7 AMEN 0 MENT NJ. 36
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SAN ONOFRE - UNIT 3 3/4 3-7a AMENDMENT NO. 36
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TABLE 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES c3 E,,
~
y FUNCTIONAL UNIT c:
RESPONSE TIME k
h 1.
Manual Reactor' Trip Not Applicable 2.
Linear Power Level - High 1 0.40 seconds
- 3.
Logarithmic Power Level - High i
5 0.45 seconds
- 4.
Presssirizer Pressure - High 5 0.90 seconds 5.
Pressurizer Pressure - Low i
$ 0.90 seconds 6.
Containment Pressure - High 5 0.90 seconds
{;
7.
Steam Generator Pressure - Low
$ 0.90 seconds
[
8.
Steam Generator Level - Low
$ 0.90 seconds 9.
Local Power Density - High Neutron Flux Power from Excore Neutron Detectors a.
5 0.68 seconds
- b.
CEA Positions
< 0.68 seconds **
c.
CEA Positions:
CEAC Penalty Factor 50.53seccnds i
j 10.
DN8R - Low i
Neutron Flux Power from Excore Neutron Detectors a.
< 0.68 seconds
- b.
CEA Positions 2 0.68 seconds **
l c.
Cold Leg Tempe'ature 50.68 seconds ##
r i
d.
Hot Leg Temperature 1 0.68 seconds ##
Primary Coolant Pump Shaf t Speed e.
1 0.68 seconds #
f.
Reactor Coolant Pressure from Pressurizer
< 0.68 seconds g.
CEA positions: CEAC Penalty Factor 50.53 seconds f
l l
i i
i
y TABLE 3.3-2 (Continued) z REACTOR _ PROTECTIVE INSTRUMENTATICi1 RESPONSE TIMES S;
FUNCTIONAL' UNIT RESPONSE TIME H
11.
Steam Generator level - High Not Applicable 12.
Reactor Protection Syst.em Logic Not Applicable 13.
Reactor Trip Breakers Not Applicable 14.
Core Protection Calculaters Not Appli. cable 13.
CEA Calculators Not Applicable 16.
Reactor Coolant flow-Low 0.9 sec 17.
Seismic-High Not Applicable
,T 18.
Loss of Load Not Applicable A
Neutron detectors are exempa from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic componcat in channel.
Response '. time shall be measurei! from the onset of a single CEA drop.
- esponse time-shall be measured using a simulated Reactor Coolant Pump coastdown.
R N Gased on a resistance temperature detector (RTD) response time cf less than or equal to 8 seconds g
where the RTD response tiae is equivalent to the time interval required for the RTD output to achieve g
r>3.2% of its total change when subjected to a step change in RTD temperature.
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- SAN ONOFRE - UNIT 3 3/4 3-9a AMEN 0 MENT NO. 36 l
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SAN ON0FRE - UNIT 3 3/4 3-9b AMEN 0 MENT NO. 36 t.
. 1......
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POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TILT - T (Continuedl T is the peak fractional tilt amplitude at the core periphery q
g is the radial normalizing factor e is the azimuthal core location 0, is the azimuthal core location of maximum tilt P
/P is the ratio of the power at a core location in the presence tilt untilt of a tilt to the power at that location with no tilt.
l 3/4.2.4 DNBR MARGIN i
The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to i
maintain an acceptable minimum DNBR throughout all anticipated operational occurrences, of which the loss of flow transient is the most limiting.
Opera-tion of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.
Either of the two core power distribuv. ion monitoring systems, the Core j
Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR.
The COLSS calculation of core power operating limit based on the minimum DNBR limit includes appropriate penalty factors which provide, with a 95/95 probability / confidence level, that the core power limit i,
calculated by COLSS (based on the minusum DNBR limit) is conservative with 3
respect to the actual core power limit.
These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, 1
engineering design factors, state parameter measurement, software algorithm modelling, computer processing, rod bow and core power measurement.
Parameters required to maintain the margin to DNB and total core power j
are also monitored by the CPCs.
In the event that the COLSS is not being used, the DNBR margin can be maintained by monitoring with any operable CPC channel so that the DNBR remains above the predetermined limit as a function of Axial Shape Index.
The above listed uncertainty penalty factors are also included in the CPCs, which assume a minimum of 20% of RATED THERMAL POWER.
The 20% RATED THERMAL POWER threshhold is due to the excore neutron flux detector system being less accurate below 20% core power.
Core. noise level at low power is too large to obtain usable detector readings.
The additional uncertainty terms taken into account in the CPCs of transient protection are removed from Figures 3.2-1 and 3.2-2 since the curves are intended to monitor the LCO only during steady state operation.
SAN ONOFRE - UNIT 3 8 3/4 2-3 AMENDMENT NO. 36 1
l
POWER DISTRIBUTION LIMITS BASES DNBR Margin (Continued)
A DNBR penalty factor has been included in the COLSS and CPC DNBR calculation to accommodate the effects of rod bow.
The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly.
Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.
Conversely, lower burnup assemblies will experience less rod bow.,
In design calculations, the penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak. A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses.
3/4.2.6 REACTOR COOLANT COLD LE3 TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses.
2.4.2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.
3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of values used in the safety analyses.
i l
(
SAN ONOFRE - UNIT 3 8 3/4 2-4 AMENDMENT N0. 36 l
T l
3/4.3 INSTRUMENTATION I
BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i
The OPERABILITY of the reactor protective and Engineered Safety Features Actuation System instrumentation and bypasses ensure that 1) the associated s
Engineered Safety Features Actuation System. action 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) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall 1
reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the 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 trip, and/or 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 i
i consistent 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 i
i 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.
1 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
)
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 frequencies are sufficient to demonstrate this capability.
The measurement of response time at the specified frequencies provides assurance that the reactor protective and ESF actuation associatec "h each channel is completed within the time limit assumed in the accident a..
/ses.
l SAN ONOFRE - UNIT 3 8 3/4 3-1 AMENDMENT NO. 36 i
i l
i
[
3/4.3 INSTRUMENTATION BASES l
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 j
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.
e l
l SAN ON0FRE - UNIT 3 B 3/4 3-la AMEN 0 MENT NO. 36 l
I INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION HONITORING ALARM INSTRUMENTATION The OPERABILITY of the radiation monitoring alarm channels ensures that
- 1) the radiation levels are continually measured in the areas served by the individual channels; 2) the alarm or automatic action 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 the recommendations of Regulatory Guide 1.97. " Instrumentation for Light-Water-Cooled Nuclear i
Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November, 1980.
3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the 1
reactor core.
3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-i bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. The instrumentation is consistent with the recommen'dations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974.
i 3/4.3.3.4.
METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological 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 and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION 4
The OPERABILITY of the remote shutdown instrumentation in Panel LO42 ensures that sufficient capability is available to permit shutdown and main-f tenance of HOT STANOBY of the facility from locations outside of the control SAN ONOFRE-UNIT 3 8 3/4 3-2 1
ADMINISTRATIVE CONTROLS
)
- 6. 4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Manager, Nuclear Training and shall meet or exceed the requirements and recommendations of Sections 5.5 of ANSI M18.1-1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the ISEG.
6.5 REVIEW AND AUDIT i
6.5.1 ONSITE REVIEW C0194ITTEE (OSRC)
FUNCTION 6.5.1.1 The Onsite Review Committee shall function to advise the Station Manager en all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The Onsite Review Committee shall be composed of the:
Chairman:
Station Manager Member:
Deputy Station Manager Member:
Manager, Operations Member:
Manager, Technical Member:
Plant Superintendent SONGS Units 2 and 3 Member:
Supervisor of I&C Member:
Manager, Health Physics Member:
Supervisor of Chemistry Member:
Manager, Maintenance Member:
Supervising Engineer (NSSS, NSSS Support, Computer, or STA)
Member:
San Diego Gas &
SeniorEngineer({}ectricRepresentative, ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the OSRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in OSRC activities at any one time.
(
BS degree in Engineering or Physical Science plus at least four years pro-fessional level experience in his field.
At least one of the four years experience shall be nuclear power plant experience.
SAN ONOFRE-UNIT 3 6-8
ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.1.4 The OSRC shall meet at least once per calendar month and as convened by the OSRC Chairman or his designated alternate.
QUORUM 6.5.1.5 The minimum quorum of the OSRC necessary for the performance of the OSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.
RESPONSIBILITIES 6.5.1.6 The Onsite Review Committee shall be responsible for:
a.
Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Nuclear Safety Group (NSG).
b.
Review of events requiring 24-hour written notification to the Commission.
Review of unit operations to detect potential nuclear safety hazards, c.
d.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Station Manager or the NSG.
Review and documentation of judgment concerning prolonged operation in e.
bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last OSRC meeting.
l l
AUTHORITY 6.5.1.7 Ine Onsite Review Committee (OSRC) shall:
a.
Render determinations in writing with regard to whether or not items considered under 6.5.1.6(a) above constitute unreviewed safety questions.
b.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Manager of Nuclear Operations and NSG of disagreement between the OSRC and the Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
I SAN ONOFRE-UNIT 3 6-9 AMENDMENT NO. 36
ADMINISTRATIVE CONTROLS ADMINISTRATIVE CONTROLS g.
PROCESS CONTROL PROGRAM implementation.*
h.
OFFSITE DOSE CALCULATION MANUAL implementation.
i.
Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15 Rev. 1, February 1979.
j.
Modification of Core Protection Calculator (CPC) Addressable Constants.
These procedures should include provisions to assure that sufficient margin is maintained in CPC Type I addressable constants to avoid ex-cessive operator interaction with the CPCs during reactor operation.
NOTE:
Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P that has been determined to be applicable to l
the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be approved by the Station Manager; or by (1) the Deputy Station Manager, (2) the Manager, Operations, (3) the Manager, Maintenance, (4) the Manager, Technical, or (5) the Manager, Health Physics as previously designated by the Station Manager; prior to implementation and shall be reviewed periodically as set forth in i
administrative procedures.
6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
c.
The change is documented, reviewed and approved by the Station Manager; or by (1) the Deputy Station Manager, (2) the Manager, i
Operations, (3) the Manager, Maintenance, (4) the Manager, Technical, or (5) the Manager, Health Physics as previously designated by the Station Manager; within 14 days of implementation.
i 6.8.4 The following programs shall be established, implemented, and maintained:
a.
Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a SAN ONOFRE-UNIT 3 6-15 AMENDMENT N0. 36
ADMINISTRATIVE CONTROLS ADMINISTRATIVE CONTROLS serious transient or accident to as low as practical levels.
The systems include the high pressure safety injection recirculation, the shutdown cooling system, the reactor coolant sampling system (post-accident sampling piping only), the containment spray system, the radioactive waste gas system (post-accident sampling return piping only) and the liquid radwaste system (post-accident sampling return piping only).
The program shall include the following:
(1) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.
"See Specification 6.13.1 SAN ONOFRE-UNIT 3 6-15a AMENDMENT NO. 36 l
_ _.