ML020720046

3. IEEE Standard 279-1971, April 5, 1972.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570, Revision 3.

7. UFSAR, Section 7.2.

8. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989.

9. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

10. Methodology for Developing Risk-Based Surveillance Programs for Safety-Related Equipment at San Onofre Nuclear Generating Station Units 2 and 3, PLG-0575, April 1992.

11. NRC Safety Evaluation Report for SONGS Unit 2 Operating License Amendment No. 150 dated February 12, 1999.

12. NRC Safety Evaluation Report for SONGS Unit 2 Operating License Amendment No. 142 dated September 25, 1998.

13. 0IC EOGýTo pfcaReport CE N*S D1167-6A, A:- mlin-atio6f AEl Pressure-Senso*r'Respo'nse Time&Testing Requirem'ents."'

SAN ONOFRE--UNIT 2 B 3.3-37a Amendment No. +/-27-G3J12ý94

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE functions, if applicable. The Surveillance verifies that REQUIREMENTS the channel responds to a measured parameter within the (continued) necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. Measurement error determination, setpoint error determination, and calibration adjustment must be performed consistent with the plant specific setpoint analysis. The channel shall be left calibrated consistent with the assumptions of the current plant specific setpoint analysis. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

SR 3.3.5.6 This Surveillance ensures that the train actuation response times are within the maximum values assumed in the safety analyses.

Response time testing acceptance criteria are included in Reference 9.

ESF RESPONSE TIME tests are conducted on a STAGGERED TEST BASIS of once every 24 months. The 24 month Frequency is consistent with the typical industry refueling cycle and is (conti nued)

SAN ONOFRE--UNIT 2 B 3.3-103a Amendment No. 15 7 1060"

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE based upon plant operating experience,which shows that REQUIREMENTS random failures of instrumentation components causing (continued) serious response time degradation, but not channel failure, are infrequent occurrences.

SR 3.3.5.7 SR 3.3.5.7 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.5.2 and SR 3.3.5.3, except SR 3.3.5.7 is performed within 120 days prior to startup and is only applicable to bypass functions. Since the Pressurizer Pressure-Low bypass is identical for both the RPS and ESFAS, this is the same Surveillance performed for the RPS in SR 3.3.1.13.

The CHANNEL FUNCTIONAL TEST for proper operation of the bypass permissives is critical during plant heatups because the bypasses may be in place prior to entering MODE 3 but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify bypass function OPERABILITY. Once the bypasses are removed, the bypasses must not fail in such a way that the associated ESFAS Function is inappropriately bypassed. This feature is verified by SR 3.3.5.2. The allowance to conduct this test once within 120 days prior to each reactor startup is based on a plant specific report based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Refs. 8 and 10).

REFERENCES 1. SONGS Units 2 and 3 UFSAR, Section 7.3.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570.

7. SONGS Units 2 and 3 UFSAR, Section 7.2.

8. CEN-327, May 1986, including Supplement 1, March 1989.

(continued)

SAN ONOFRE--UNIT 2 B 3.3-103b Amendment No. 157 --

ESFAS Instrumentation B 3.3.5 BASES (continued)

REFERENCES 9. Licensee Controlled Specification 3.3.100, "RPS/ESFAS (Continued) Response Times."

10. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

11. Report NSG 98-007, "Time Limit for RAS or EFAS Channel in Trip," April 17, 1998.

12. .CEOG Topical Report CE NPSD-.1167-A, "Elimination of Pressure Sensor Response ,Time iTest!ing. Requirements." I SAN ONOFRE--UNIT 2 B 3.3-103c Amendment No.

Definitions 1.1 1.1 Definitions CORE ALTERATION within the reactor vessel with the vessel head (continued) removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.7.1.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP-30, Supplement to Part 1, pages 192-212, Tables titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

E- AVERAGE F shall be the average (weighted in proportion DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its ESF TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of (continued)

SAN ONOFRE--UNIT 3 1.1-3 Amendment No. 437

Definitions 1.1 1.1 Definitions ENGINEERED SAFETY measurement, response time may be verified for FEATURE (ESF) RESPONSE selected components provided that the components TIME (Continued) and methodoloqy for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System.

b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(conti nued)

SAN ONOFRE--UNIT 3 1.1-4 Amendment No.+&

Definitions 1.1 1.1 Definitions OPERABLE- OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Test Program of the SONGS Units 2 and 3 UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3438 MWt.

REACTOR PROTECTIVE The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. n,.`1eu i of:meas-uriement;response tlme m beeIfied for'se"ected ci~mpnts provided that the-components and methodol oqy .for,,

verification have been previ~ously rev'iewed ýand appypved b4ythe .NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

(conti nued)

SAN ONOFRE--UNIT 3 1.1-5 Amendment No. 44+

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) a. All full length CEAs (shutdown and regulating)

(continued) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CEAs verified fully inserted by two independent means, it is not necessary to account for a stuck CEA in the SDM calculation.

b. In MODES I and 2, the fuel and moderator temperatures are changed to the corrected hot zero power condition.

c. There is no change in part length CEA position.

With any CEAs not capable of being fully inserted, the reactivity worth of these CEAs must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

SAN ONOFRE--UNIT 3 1.1-6 Amendment No. +4-I-

RPS Instrumentation -Operating B 3.3.1 BASES (continued)

SURVEILLANCE REQUIREMENTS (continued) This SR ensures that the RPS RESPONSE TIMES are verified to be less than or equal to the maximum values assumed in the safety analysis. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the RTCBs open. Response times are conducted on an 24 month STAGGERED TEST BASIS. This results in the interval between successive surveillances of a given channel of n x 24 months, where n is the number of channels in the function. The Frequency of 24 months is based upon operating experience, which has shown that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. Also, response times cannot be determined at power, since equipment operation is required.-Testg may be performed in e,,measurement or in overlapping segments, with verifieatien that all eamponents are tested.

Rei'p6o ' e,_t im~e"`th__a`VIieb -v'erff i..ed- by,:-any' s-6"rfies'ýof -se~quential, overta'ppin'q'ý6iý'-ttal~ch n'nel measurements, includin allocated sehsoiý response time, ýý'ch.'ha-h'-ep~ tim isfe:~1ocainfr s0enso Ir-respons-eI1jesýmaybe obandf6,r.d :ts results, -vend'o'.-tst'dt :o r verdeignee#, 1 rlspeSPCIfications. .Topical xeport E NP$D*~t6~A

~iE~PiLii i~toii~bf Tres §'ure -, Se soR b eTjm A Note is added to indicate that the neutron detectors are excluded from RPS RESPONSE TIME testing because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in leakage of neutrons with core burnup are compensated for by performing the daily calorimetric calibration (SR 3.3.1.4).

(continued)

SAN ONOFRE--UNIT 3 B 3.3-37 Amendment No. -116 -3/2ýG 4

RPS Instrumentation -Operating B 3.3.1 BASES (continued)

REFERENCES 1. 10 CFR 20.

3. IEEE Standard 279-1971, April 5, 1972.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570, Revision 3.

7. UFSAR, Section 7.2.

8. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989.

9. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

10. Methodology for Developing Risk-Based Surveillance Programs for Safety-Related Equipment at San Onofre Nuclear Generating Station Units 2 and 3, PLG-0575, April 1992.

11. NRC Safety Evaluation Report for SONGS Unit 3 Operating License Amendment No. 142 dated February 12, 1999.

12. NRC Safety Evaluation Report for SONGS Unit 3 Operating License Amendment No. 136 dated November 23, 1998.

i3."CEOGTop*ai- Report tCE .NPSD, 1167"-A,' "El iminat ion of P.essuree, orResponse:jime., esting. Requirements.;,

SAN ONOFRE--UNIT 3 B 3.3-37a Amendment No. 16- 03-21

ESFAS Instrumentation BASES (continued) B 3.3.5 SURVEILLANCE functions, if applicable. The Surveillance verifies that REQUIREMENTS the channel responds to a measured parameter within the (continued) necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. Measurement error determination, setpoint error determination, and calibration adjustment must be performed consistent with the plant specific setpoint analysis. The channel shall be left calibrated consistent with the assumptions of the current plant specific setpoint analysis. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

SR 3.3.5.6 This Surveillance ensures that the train actuation response times are within the maximum values assumed in the safety analyses.

Response time testing acceptance criteria are included in Reference 9.

ESF RESPONSE TIME tests are conducted on a STAGGERED TEST BASIS of once every 24 months. The 24 month Frequency is consistent with the typical industry refueling cycle and is (continued)-

SAN ONOFRE--UNIT 3 B 3.3-103a Amendment No. 4006/99

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE based upon plant operating experience, which shows that REQUIREMENTS random failures of instrumentation components causing (continued) serious response time degradation, but not channel failure, are infrequent occurrences.

SR 3.3.5.7 SR 3.3.5.7 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.5.2 and SR 3.3.5.3, except SR 3.3.5.7 is performed within 120 days prior to startup and is only applicable to bypass functions. Since the Pressurizer Pressure-Low bypass is identical for both the RPS and ESFAS, this is the same Surveillance performed for the RPS in SR 3.3.1.13.

The CHANNEL FUNCTIONAL TEST for proper operation of the bypass permissives is critical during plant heatups because the bypasses may be in place prior to entering MODE 3 but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify bypass function OPERABILITY. Once the bypasses are removed, the bypasses must not fail in such a way that the associated ESFAS Function is inappropriately bypassed. This feature is verified by SR 3.3.5.2. The allowance to conduct this test once within 120 days prior to each reactor startup is based on a plant specific report based on the reliability analysis presented in topical report CEN-327, "RPS/ ESFAS Extended Test Interval Evaluation" (Refs. 8 and 10).

REFERENCES 1. SONGS Units 2 and 3 UFSAR, Section 7.3.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570.

7. SONGS Units 2 and 3 UFSAR, Section 7.2.

8. CEN-327, May 1986, including Supplement 1, March 1989.

(continued)

SAN ONOFRE--UNIT 3 B 3.3-103b Amendment No. 148 101067t9

ESFAS Instrumentation B 3.3.5 BASES (continued)

REFERENCES 9. Licensee Controlled Specification 3.3.10, "RPS/ESFAS (Continued) Response Times."

10. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

11. Report NSG 98-007, "Time Limit for RAS or EFAS Channel in Trip," April 17, 1998.

12. CEOG Topi ca*l Report CEt NPSD-1167-A, imi nat i on of El*

Pressur-e-'$ensor Response T1Time Testfing Requr ements."s SAN ONOFRE--UNIT 3 B 3.3-103C. Amendment No.

ATTACHMENT 4 SAN ONOFRE NUCLEAR GENERATING STATION PCN-539 RETYPED TECHNICAL SPECIFICATION PAGES AND BASES (for information only) PAGES

Definitions 1.1 1.1 Definitions CORE ALTERATION within the reactor vessel with the vessel head (continued) removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.7.1.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP-30, Supplement to Part 1, pages 192-212, Tables titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

E- AVERAGE E shall be the average (weighted in proportion DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its ESF TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of (continued)

SAN ONOFRE--UNIT 2 1.1-3 Amendment No.

Definitions 1.1 1.1 Definitions ENGINEERED SAFETY measurement, response time may be verified for FEATURE (ESF) RESPONSE selected components provided that the components TIME (Continued) and methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System.

b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(continued)

SAN ONOFRE--UNIT 2 1.1-4 Amendment No.

Definitions 1.1 1.1 Definitions OPERABLE- OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Test Program of the SONGS Units 2 and 3 UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3438 MWt.

REACTOR PROTECTIVE The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

(continued)

SAN ONOFRE--UNIT 2 1.1-5 Amendment No.

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) a. All full length CEAs (shutdown and regulating)

(continued) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CEAs verified fully inserted by two independent means, it is not necessary to account for a stuck CEA in the SDM calculation.

b. In MODES I and 2, the fuel and moderator temperatures are changed to the corrected hot zero power condition.

c. There is no change in part length CEA position.

With any CEAs not capable of being fully inserted, the reactivity worth of these CEAs must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

SAN ONOFRE--UNIT 2 1.1-6 Amendment No.

RPS Instrumentation-Operating B 3.3.1 BASES (continued)

SURVEILLANCE SR 3.3.1.13 REQUIREMENTS (continued) This SR ensures that the RPS RESPONSE TIMES are verified to be less than or equal to the maximum values assumed in the safety analysis. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the RTCBs open. Response times are conducted on an 24 month STAGGERED TEST BASIS. This results in the interval between successive surveillances of a given channel of n x 24 months, where n is the number of channels in the function. The Frequency of 24 months is based upon operating experience, which has shown that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. Also, response times cannot be determined at power, since equipment operation is required.

Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specifications. Topical Report CE NPSD-1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements," Reference 13, provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the Topical Report. Response time verification for other sensor types must be demonstrated by test. The allocation of sensor response times must be verified prior to placing a new component in operation and reverified after maintenance that may adversely affect the sensor response time. The replaced or repaired sensor must be tested to determine actual response time.

A Note is added to indicate that the neutron detectors are excluded from RPS RESPONSE TIME testing because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in leakage of neutrons with core burnup are compensated for by performing the daily calorimetric calibration (SR 3.3.1.4).

(continued)

SAN ONOFRE--UNIT 2 B 3.3-37 Amendment No.

RPS Instrumentation-Operating B 3.3.1 BASES (continued)

REFERENCES 1. 10 CFR 20.

3. IEEE Standard 279-1971, April 5, 1972.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570, Revision 3.

7. UFSAR, Section 7.2.

8. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989.

9. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

10. Methodology for Developing Risk-Based Surveillance Programs for Safety-Related Equipment at San Onofre Nuclear Generating Station Units 2 and 3, PLG-0575, April 1992.

11. NRC Safety Evaluation Report for SONGS Unit 2 Operating License Amendment No. 150 dated February 12, 1999.

12. NRC Safety Evaluation Report for SONGS Unit 2 Operating License Amendment No. 142 dated September 25, 1998.

13. CEOG Topical Report CE NPSD-1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements."

SAN ONOFRE--UNIT 2 B 3.3-37a Amendment No.

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE functions, if applicable. The Surveillance verifies that REQUIREMENTS the channel responds to a measured parameter within the (continued) necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. Measurement error determination, setpoint error determination, and calibration adjustment must be performed consistent with the plant specific setpoint analysis. The channel shall be left calibrated consistent with the assumptions of the current plant specific setpoint analysis. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

SR 3.3.5.6 This Surveillance ensures that the train actuation response times are within the maximum values assumed in the safety analyses.

Response time testing acceptance criteria are included in Reference 9.

Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specifications. Topical Report CE NPSD 1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements," Reference 12, provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the Topical Report. Response time verification for other sensor types must be demonstrated by test. The allocation of sensor response times must be verified prior to placing a new component in operation and reverified after maintenance that may adversely affect the sensor response time. The replaced or repaired sensor must be tested to determine actual response time.

ESF RESPONSE TIME tests are conducted on a STAGGERED TEST BASIS of once every 24 months. The 24 month Frequency is consistent with the typical industry refueling cycle and is (continued)

SAN ONOFRE--UNIT 2 B 3.3-103a Amendment No.

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE based upon plant operating experience,which shows that REQUIREMENTS random failures of instrumentation components causing (continued) serious response time degradation, but not channel failure, are infrequent occurrences.

SR 3.3.5.7 SR 3.3.5.7 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.5.2 and SR 3.3.5.3, except SR 3.3.5.7 is performed within 120 days prior to startup and is only applicable to bypass functions. Since the Pressurizer Pressure-Low bypass is identical for both the RPS and ESFAS, this is the same Surveillance performed for the RPS in SR 3.3.1.13.

The CHANNEL FUNCTIONAL TEST for proper operation of the bypass permissives is critical during plant heatups because the bypasses may be in place prior to entering MODE 3 but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify bypass function OPERABILITY. Once the bypasses are removed, the bypasses must not fail in such a way that the associated ESFAS Function is inappropriately bypassed. This feature is verified by SR 3.3.5.2. The allowance to conduct this test once within 120 days prior to each reactor startup is based on a plant specific report based on the reliability analysis presented in topical report CEN-327, "RPS ESFAS Extended Test Interval Evaluation" (Refs. 8 and 10).

REFERENCES 1. SONGS Units 2 and 3 UFSAR, Section 7.3.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570.

7. SONGS Units 2 and 3 UFSAR, Section 7.2.

8. CEN-327, May 1986, including Supplement 1, March 1989.

(continued)

SAN ONOFRE--UNIT 2 B 3.3-103b Amendment No.

ESFAS Instrumentation B 3.3.5 BASES (continued)

REFERENCES 9. Licensee Controlled Specification 3.3.100, "RPS/ESFAS (Continued) Response Times."

10. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

11. Report NSG 98-007, "Time Limit for RAS or EFAS Channel in Trip," April 17, 1998.

12. CEOG Topical Report CE NPSD-1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements."

SAN ONOFRE--UNIT 2 B 3.3-103c Amendment No.

Definitions 1.1 1.1 Definitions CORE ALTERATION within the reactor vessel with the vessel head (continued) removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.7.1.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP-30, Supplement to Part 1, pages 192-212, Tables titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

r - AVERAGE E shall be the average (weighted in proportion DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its ESF TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of (conti nued)

SAN ONOFRE--UNIT 3 1.1-3 Amendment No.

Definitions 1.1 1.1 Definitions ENGINEERED SAFETY measurement, response time may be verified for FEATURE (ESF) RESPONSE selected components provided that the components TIME (Continued) and methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System.

b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

(conti nued)

SAN ONOFRE--UNIT 3 1.1-4 Amendment No.

Definitions 1.1 1.1 Definitions OPERABLE- OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Test Program of the SONGS Units 2 and 3 UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3438 MWt.

REACTOR PROTECTIVE The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

(conti nued)

SAN ONOFRE--UNIT 3 1.1-5 Amendment No.

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) a. All full length CEAs (shutdown and regulating)

(continued) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all CEAs verified fully inserted by two independent means, it is not necessary to account for a stuck CEA in the SDM calculation.

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the corrected hot zero power condition.

c. There is no change in part length CEA position.

With any CEAs not capable of being fully inserted, the reactivity worth of these CEAs must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

SAN ONOFRE--UNIT 3 1.1-6 Amendment No.

RPS Instrumentation-Operating B 3.3.1 BASES (continued)

SURVEILLANCE SR 3.3.1.13 REQUIREMENTS (continued) This SR ensures that the RPS RESPONSE TIMES are verified to be less than or equal to the maximum values assumed in the safety analysis. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the RTCBs open. Response times are conducted on an 24 month STAGGERED TEST BASIS. This results in the interval between successive surveillances of a given channel of n x 24 months, where n is the number of channels in the function. The Frequency of 24 months is based upon operating experience, which has shown that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. Also, response times cannot be determined at power, since equipment operation is required.

Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specifications. Topical Report CE NPSD-1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements," Reference 13, provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the Topical Report. Response time verification for other sensor types must be demonstrated by test. The allocation of sensor response times must be verified prior to placing a new component in operation and reverified after maintenance that may adversely affect the sensor response time. The replaced or repaired sensor must be tested to determine actual response time.

A Note is added to indicate that the neutron detectors are excluded from RPS RESPONSE TIME testing because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in leakage of neutrons with core burnup are compensated for by performing the daily calorimetric calibration (SR 3.3.1.4).

(continued)

SAN ONOFRE--UNIT 3 B 3.3-37 Amendment No.

RPS Instrumentation-Operating B 3.3.1 BASES (continued)

REFERENCES 1. 10 CFR 20.

3. IEEE Standard 279-1971, April 5, 1972.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.

6. PPS Setpoint Calculation CE-NPSD-570, Revision 3.

7. UFSAR, Section 7.2.

8. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989.

9. RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing at SONGS Units 2 and 3, Calculation Number 09/010-AS93-C-002, November 1993.

10. Methodology for Developing Risk-Based Surveillance Programs for Safety-Related Equipment at San Onofre Nuclear Generating Station Units 2 and 3, PLG-0575, April 1992.

11. NRC Safety Evaluation Report for SONGS Unit 3 Operating License Amendment No. 142 dated February 12, 1999.

12. NRC Safety Evaluation Report for SONGS Unit 3 Operating License Amendment No. 136 dated November 23, 1998.

13. CEOG Topical Report CE NPSD-1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements."

SAN ONOFRE--UNIT 3 B 3.3-37a Amendment No.

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE functions, if applicable. The Surveillance verifies that REQUIREMENTS the channel responds to a measured parameter within the (continued) necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. Measurement error determination, setpoint error determination, and calibration adjustment must be performed consistent with the plant specific setpoint analysis. The channel shall be left calibrated consistent with the assumptions of the current plant specific setpoint analysis. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

SR 3.3.5.6 This Surveillance ensures that the train actuation response times are within the maximum values assumed in the safety analyses.

Response time testing acceptance criteria are included in Reference 9.

Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified. Allocations for sensor response times may be obtained from records of test results, vendor test data, or vendor engineering specifications. Topical Report CE NPSD 1167-A, "Elimination of Pressure Sensor Response Time Testing Requirements," Reference 12, provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the Topical Report. Response time verification for other sensor types must be demonstrated by test. The allocation of sensor response times must be verified prior to placing a new component in operation and reverified after maintenance that may adversely affect the sensor response time. The replaced or repaired sensor must be tested to determine actual response time.

ESF RESPONSE TIME tests are conducted on a STAGGERED TEST BASIS of once every 24 months. The 24 month Frequency is consistent with the typical industry refueling cycle and is (conti nued)

SAN ONOFRE--UNIT 3 B 3.3-103a Amendment No.

ESFAS Instrumentation B 3.3.5 BASES (continued)

SURVEILLANCE based upon plant operating experience, which shows that REQUIREMENTS random failures of instrumentation components causing (continued) serious response time degradation, but not channel failure, are infrequent occurrences.

SR 3.3.5.7 SR 3.3.5.7 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.5.2 and SR 3.3.5.3, except SR 3.3.5.7 is performed within 120 days prior to startup and is only applicable to bypass functions. Since the Pressurizer Pressure-Low bypass is identical for both the RPS and ESFAS, this is the same Surveillance performed for the RPS in SR 3.3.1.13.

The CHANNEL FUNCTIONAL TEST for proper operation of the bypass permissives is critical during plant heatups because the bypasses may be in place prior to entering MODE 3 but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify bypass function OPERABILITY. Once the bypasses are removed, the bypasses must not fail in such a way that the associated ESFAS Function is inappropriately bypassed. This feature is verified by SR 3.3.5.2. The allowance to conduct this test once within 120 days prior to each reactor startup is based on a plant specific report based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Refs. 8 and 10).

REFERENCES 1. SONGS Units 2 and 3 UFSAR, Section 7.3.

4. SONGS Units 2 and 3 UFSAR, Chapter 15.