ML13331A438

Supplemental Amend Application 161,consisting of Proposed Change 183,Rev 1 to License DPR-13.Amend Adds Limiting Conditions for Operation & Surveillance Requirements for Reactor Protection Sys Components

ML13331A438
Person / Time
Site:	San Onofre
Issue date:	06/15/1990
From:	Ray H SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML13331A437	List: ML13331A436 ML13331A438 ML13331A439
References
GL-85-09, GL-85-9, NUDOCS 9006190079
Download: ML13331A438 (21)

Text

BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON

)

COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY

)

DOCKET NO. 50-206 for a Class 104(b) License to Acquire,

)

Possess, and Use a Utilization Facility as

)

Supplemental Part of Unit No. 1 of the San Onofre Nuclear

)

Amendment Generating Station

)

Application No. 161 SOUTHERN CALIFORNIA EDISON COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY, pursuant to 10 CFR 50.90, hereby submit Supplemental Amendment Application No. 161.

This supplemental amendment consists of Proposed Change No.

183 Revision 1 to Provisional Operating License No. DPR-13.

Proposed Change No. 183 Revision 1 revises Specification 3.5.1, "Reactor Trip System Instrumentation",

and Specification 4.1-1, " Operational Safety Items".

Proposed Change No. 183 Revision 1 is a request to add limiting conditions for operation and surveillance requirements for Reactor Protection System components into the technical specifications that are currently performed by procedure only.

In addition requirements for the testing of the reactor trip breakers undervoltage and shunt trip attachments are incorporated in response to Generic Letter 85-09, "Technical Specifications for Generic Letter 83-28, Item 4.3".

Proposed Change No.

183 Revision 1 supercedes and rescinds our request to incorporate Westinghouse Owners Group recommendations to increase surveillance intervals for Reactor Protection System components.

9006190079 900615' PDR ADOCK 0500:C206 P

PDIC

In the event of conflict, the information in Supplemental Amendment Application No. 161 Proposed Change No. 183 Revision 1 supersedes the information previously submitted.

Based on the significant hazards analysis provided in the Description and Significant Hazards Consideration Analysis of Proposed Change No.

183, it is concluded that (1) the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.

-3 Subscribed on this

/5 day of

/909, 1990.

Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By: Hald B.

Ray Senior Vice Pr s ent Subscribed and sworn to before me this

/-day of

/FFIC.

YOOMAL y

VOMAR V. CLEARY 3 day o

Notery Pt*llo-CaifomIs My Comm. Exp. May8 192 Not y Public in and for tMi Sta e of California James A. Beoletto Attorney for Southern California Edison Company By:

Jame A.

eolettoF"

0 0

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN CALIFORNIA

)

EDISON COMPANY and SAN DIEGO GAS &

)

Docket No. 50-206 ELECTRIC COMPANY (San Onofre Nuclear

)

Generating Station, Unit No. 1)

)

CERTIFICATE OF SERVICE I hereby certify that a copy of Supplemental Amendment Application No. 161 was served on the following by deposit in the United States Mail, postage prepaid, on the 15th day of June

, 1990.

Benjamin H. Vogler, Esq.

Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C.

20555 David R. Pigott, Esq.

Samuel B. Casey, Esq.

Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P.O. Box 60860, Terminal Annex Los Angeles, California 90060 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102 C. J. Craig Manager U.S. Nuclear Projects I ESSD Westinghouse Electric Corporation Post Office Box 355 Pittsburgh, Pennsylvania 15230

-2 A. I. Gaede 23222 Cheswald Drive Laguna Niguel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.

20555 aes A. Beoletto-

Revision 1 DESCRIPTION AND SAFETY ANALYSIS FOR SUPPLEMENTAL CHANGE TO PROPOSED CHANGE NO. 183 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-13 This is a request to revise Section 3.5.1, "Reactor Trip System Instrumentation," and Section 4.1.1, "Operational Safety Items," of Appendix A Technical Specifications for San Onofre Nuclear Generating Station Unit 1.

DESCRIPTION Amendment Application No. 161 was submitted to the NRC on December 29, 1988.

This supplement to Amendment Application No. 161 consists of a revision to Proposed Change No. 183 (PCN 183). This Proposed Change to San Onofre Nuclear Generating Station Unit 1 Appendix A Technical Specification Table 3.5.1-1, "Reactor Trip System Instrumentation", and Table 4.1-1, "Reactor Trip System Instrumentation Surveillance Requirements", is provided in resolution of Systematic Evaluation Program (SEP) Topic VI-10.A, "Testing of Reactor Trip System and Engineered Safety Features".

In addition the PCN incorporates requirements for the testing of the reactor trip breakers undervoltage and shunt trip attachments in response to Generic Letter 85-09, "Technical Specifications for Generic Letter 83-28, Item 4.3". The revised PCN supercedes and rescinds our request to incorporate Westinghouse Owners Group (WOG) recommendations to increase surveillance intervals for Reactor Protection System components as documented in WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System".

BACKGROUND Application of SEP Topic VI-10.A In the San Onofre Unit 1 Technical Specifications, the channel response time from channel trip to deenergization of the scram relay is not tested. However the reactor trip breakers are tested and the rod drop time is measured and must be within specified limits (in the Technical Specifications).

These aspects are also covered by plant procedures.

Several channels were not included in the Technical Specifications to be checked, tested, or calibrated. However, station procedures establish testing requirements for these channels at frequencies consistent with Standard Technical Specification requirements.

The NRC staff performed a limited PRA of this issue for SONGS 1 to estimate the improvement if response-time testing of the reactor protection system (RPS) were required. The results of this assessment indicated that response time testing has low safety significance. This occurs because response-time testing is concerned with events on the order of seconds and PRAs have shown that response times of minutes are sufficient, for RPS actuation, to ensure the success of the reactivity control function in time to allow other safety

-2 systems to act to prevent core melt.

Functional tests, such as those currently performed at SONGS 1, are sufficient to demonstrate functioning on the order of minutes. Therefore, the NRC staff determined that response-time testing of the RPS is not required. However, the RPS testing currently covered by plant procedures should be incorporated into the Technical Specifications because of the safety significance of this system. By letter dated March 30, 1984, SCE agreed to propose a Technical Specification change to incorporate channel testing, checking, and calibration requirements for those RPS channels currently specified by procedure only.

Application of WOG Guidelines While reviewing PCN 183, the NRC discovered several discrepancies in the submittal.

By letter dated October 31, 1989, the NRC requested that we provide additional information to clarify these discrepancies. In response to this request, SCE initiated a detailed review of the PCN. We discovered further discrepancies associated with the use of WOG recommendations to extend the RPS surveillance intervals. The WOG guidelines were based on the design of a Westinghouse generic plant that conforms to the Westinghouse Standard Technical Specifications (STS).

SONGS 1 is not consistent with the Westinghouse generic design or the STS.

As an example, these plants have bypass switching capabilities that are not available at SONGS 1.

The NRC stated that the acceptability of a licensees' response is contingent on confirmation of the capability to perform the RPS analog testing in bypass without the use of lifted leads or temporary jumpers. Since no provisions have been made by SCE to install the hardware to perform the testing in the bypass condition, we need to further evaluate the applicability of the WOG guidelines for the SONGS 1 design. For this reason we have decided to withdraw our request to extend the RPS instrumentation surveillance intervals.

If we determine this application is practicable, then we will submit a separate amendment to incorporate the appropriate change.

Application of Generic Letter 85-09 Item 4.3 of Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events", established the requirement for the automatic actuation of the shunt trip attachment for-Westinghouse plants.

Based on this requirement Westinghouse developed a generic design modification for the reactor trip breakers (RTB).

In the NRC staff's evaluation of the Westinghouse generic design modification, the staff concluded that Technical Specification changes should be proposed by licensees to explicitly require independent testing of the undervoltage and shunt trip attachments during power operation and independent testing of the control room manual switch contacts during each refueling outage. The staff concluded that these tests are necessary to ensure reliable reactor breaker operation.

-3 SONGS 1 is not consistent with the Westinghouse generic design modification for reactor trip breakers. Without bypass breakers, no capability exists at SONGS 1 to test the reactor trip breakers while the unit is operating. In response to Generic Letter 85-09, the proposed refueling outage test frequency is based on the SONGS 1 RTB configuration of two breakers in series with no bypass breakers. By letter dated September 16, 1988 the NRC issued a Safety Evaluation in which they determined that our RTB control scheme is fail-safe and reliable, and that the design changes that would be required to perform on-line RTB testing will not improve overall reliability, therefore, such on line testing of the RTBs is not required.

DESCRIPTION OF CHANGES Revisions to San Onofre Unit 1 Reactor Trip System Instrumentation Technical Specifications are proposed as follows:

1.

Incorporate channel testing, checking and calibration requirements currently specified only by procedure (SEP Topic VI-10.A) for the Reactor Trip Breakers and Sequencer Input to Reactor Trip into Table 3.5.1-1 and Table 4.1.1.

2.

Incorporate requirements to independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers (Generic Letter 85-09) into Table 3.5.1-1 and Table 4.1.1.

Each of the above changes is discussed in detail in the following sections.

The proposed technical specifications that incorporate these changes are provided in Attachment 2.

I.

The proposed change to Technical Specification Table 3.5.1-1, "Reactor Trip System Instrumentation", would revise the instrumentation and control requirements to incorporate Limiting Conditions for Operation as described below.

1.

Reactor Trip Breakers In conjunction with the resolution to SEP Topic VI-10.A, an LCO for this parameter will be incorporated into the technical specifications. This parameter is currently controlled by plant procedures. The proposed LCO mode applicability requirements are consistent with the Westinghouse STS. The basis for the corresponding ACTION statements is provided below.

ACTION 30 has been added to Table 3.5.1-1 in order to establish appropriate controls for inoperability of the reactor trip breakers (RTBs). This action is similar to ACTION 29 with the exception that no bypass capability is provided for the SONGS 1

0 0

-4 RTBs. The configuration of the RTBs is two breakers in series with no bypass breakers. Therefore, should one breaker become inoperable, the plant must be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. During this time, the remaining OPERABLE breaker will be available to trip the plant if necessary. Once HOT STANDBY has been achieved, ACTION 7 will be imposed. ACTION 7 allows 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore the inoperable breaker to OPERABLE status or open the RTBs within the following hour.

2.

Sequencer Input to Reactor Trip In conjunction with the resolution to SEP Topic VI-10.A, an LCO for this parameter will be incorporated into the technical specifications. This parameter is currently controlled by plant procedures. The configuration for this reactor trip signal is as follows. SONGS 1 utilizes two sequencers to automatically start emergency loads in the event of a safety injection actuation signal or automatically sequence emergency loads onto the emergency diesel generators in the event of a coincident safety injection signal and loss of offsite power. Each sequencer has two subchannels. These subchannels each send a signal, initiated by SIS or LOP, to a logic gate. When positive signals are sent from both subchannels, a reactor trip signal is initiated. This process is the same for both sequencers.

The proposed channel operability requirements stipulate that both sequencers be OPERABLE with one sequencer required to send a trip signal. The operability of the sequencers includes the individual subchannels. ACTION 29 will be applied to establish appropriate controls for the inoperability of a sequencer channel. A footnote has also been added to require compliance with the provisions of Specification 3.5.5 for any portion of this channel required to be OPERABLE by Specification 3.5.5. This footnote is added to ensure compliance with the Containment Isolation function of this channel.

II. The proposed change to Technical Specification Table 4.1.1, "Reactor Trip System Instrumentation Surveillance Requirements", would revise the RPS instrumentation and control surveillance requirements as described below.

1.

Reactor Trip Breakers As part of the resolution of SEP Topic VI-10.A, this channel, which is currently tested by procedure only, will now be included in the technical specifications. The proposed refueling outage test frequency is based on the SONGS 1 RTB configuration of two breakers in series with no bypass breakers. Without bypass breakers, no capability exists to test the breakers while the unit is operating. Footnote 8 will be included to require that the surveillance tests of the Reactor Trip Breakers independently

-5 verify the OPERABILITY of the undervoltage and shunt trip attachments. This footnote is incorporated in response to Generic Letter 85-09.

2.

Sequencer Input to Reactor Trip As part of the resolution of SEP Topic VI-10.A, this channel, which is currently tested by procedure only, will be incorporated into the technical specifications. This channel is essentially the same as the STS Safety Injection Input from ESF channel.

The proposed test frequency is consistent with the STS for these channels.

Footnote 9 has been added to require compliance with the provisions of Specification 4.1.4 for any portion of this channel required to be OPERABLE by Specification 3.5.5. This footnote is added to ensure compliance with the Containment Isolation function of this channel.

3.

Manual Reactor Trip Footnote 7 has been included to require that the TRIP ACTUATING DEVICE OPERATIONAL TEST independently verify the operability of the undervoltage and shunt trip circuits for the Manual Reactor Trip function. This footnote is incorporated in response to NRC Generic letter 85-09. It is noted that the WOG requirement to verify the operability of bypass breakers has not been included since SONGS 1 does not utilize bypass breakers.

EXISTING TECHNICAL SPECIFICATIONS The existing Technical Specifications are provided in Attachment 1.

PROPOSED TECHNICAL SPECIFICATIONS The proposed Technical Specifications are provided in Attachment 2. Change bars are used to illustrate the proposed revisions.

SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS As required by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that the proposed license amendment to implement technical specifications associated with Reactor Protection System instrumentation at SONGS 1 represents a no significant hazards consideration. In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed amendment was analyzed using the following standards and found not to:

1) involve a significant increase in the probability or consequences for an accident

-6 previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change does not result in an increase in the probability or consequences of an accident previously evaluated. The proposed change does not involve a modification to the plant design.

Implementation of the proposed change does not alter the manner in which the RPS is operated and maintained. The proposed change will incorporate into the Technical Specifications refueling interval RPS surveillance testing currently covered by procedure.

This proposed change is provided, as requested by the NRC, in resolution to SEP Topic VI-10.A, "Testing of Reactor Trip System and Engineered Safety Features". The proposed Limiting Conditions for Operation incorporated by this change is consistent with the Westinghouse Standard Technical Specifications.

This change incorporates the requirements for the refueling interval surveillance testing of the reactor trip breaker undervoltage and shunt trip attachments. This testing is also performed as part of procedures. This proposed change is provided, as requested by the NRC, in response to Generic Letter 85-09, "Technical Specifications for Generic Letter 83-28, Item 4.3".

Therefore, operation of the facility in accordance with this proposed change will not increase the probability or consequences of and accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change does not result in a new or different kind of accident from any accident previously evaluated. Implementation of the proposed change does not alter the manner in which the RPS is operated and maintained. The proposed change does not affect the manner in which the RPS provides plant protection. The proposed RPS testing has been previously reviewed and found to be acceptable.

Therefore, implementation of this proposed change for testing does not create the possibility of a new or different kind of accident

-7 from any previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change does not alter the manner in which the RPS is operated or maintained. The proposed change will incorporate into the Technical Specifications refueling interval RPS surveillance testing currently covered by procedure. The proposed Limiting Conditions for Operation are consistent with the Westinghouse Standard Technical Specifications and do not involve reductions in the safety limits or the limiting safety system setpoints for the RPS. The proposed refueling interval surveillance testing of the RPS instrumentation will be performed during a plant mode in which the components tested will not be required to perform a safety function.

Therefore operation of the facility in accordance with this proposed change will not involve a reduction in a margin of safety.

SAFETY AND SIGNIFICANT HAZARDS DETERMINATION Based on the safety analysis, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.

ATTACHMENT 1 EXISTING TECHNICAL SPECIFICATIONS 3.5 INSTRUMENTATTON AND CONTRgL 3.5.1 REACTOR TRIP SYSTEM TuNSTRUMENTATION APPLICABILITY:

As shown in Table 3.5.1-1.

OBJECTIVE:

To delineate the conditions of the Plant instrumentation and safety circuits necessary to ensure reactor safety.

SPECIFICATION:

As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3M5.1-1shall be OPERA8LE.

ACTION:

As shown in Table 3.5.1-1.

During plant operations, the complete instrumentation systems will normally be in service.(l) Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established llmits.(Z) Safety is not compromised, however. by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design.(1)(3) This Standard outlines limiting conditions for operation necessary to preserve the effectiveness of the reactor control and protection system when any one or more of the channels is out of service.

al S

REFERNCES:

(1) Final Engineering Report and Safety Analysis, Section 6.

(2) Final Engineering Report and Safety Analysis,

.Section 6.2.

(3) NIS Safety Review Report, April 1988 SAN ONOFRE - UNIT 1 3.5-1 AMENDMENT NO: 83, 117, 130

TALE 3.5.1-1 (A

REACTOR TRIP SYSTEM INSIRUMENTATION NINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNTION UNII OF CHANNELS TO TRIP OPERABLE MODES ACTION awl

1. Manual Reactor Trip 2

1 2

1, 2 I

2 1

2 3

4*, 5 1

2. Power Range. Neutron Flux.

4 2

3

1. 2 21 Overpower Trip

3. Power Range. Neutron flux.

4 I"

4

1. 2 280 Dropped Rod Rod Stop

4. Intermediate Range. Neutron 2

I 2

1l.

2**

3 Flux

5. Source Range. Neutron Flux A. Startup 2

1" 2

200 4

B. Shutdown 2

2

3. 4'. 5*

7 C. Shutdown 2

0 1

3. 4. and 5 5

6.

NIS Colncidentor Logic 2

1 2

1. 2 29 V3* 4%, 50 7

1. Pressurizer Variable 3

2 2

Idle 60 Low Pressure

8. Pressurizer fixed High 3

2 2

1. 2 60 Pressure

9. Pressurizer high Level 3

2 2

1 60 C

00 on 0

C CD*

TABLE 3.5.1-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION o

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT Of CMANNELs To RIP OPERABLE hoofs ACTION

10. Reactor Coolant Flow C

A. Single Loop 1/loop I/loop in any I/loop in each I

i (Above 50% of Full Power) operating loop operating loop

-4

8. Two Loops I/loop I/loop in two I/loop in each lo##

6#

(Below 50% of Full Power) operating loops operating loop II.

Steam/Feecheater Flow Mismatch 3

2 2

100000 60

12.

Turbine Trip-Low Fluid Oil Pressure 3

2 2

1I##

6f

13.

Reactor Coolant Pump Breaker Position A. Single Loop I/loop I/loop in any I/loop in each I

60 (Above 50% of Full Power) operating loop operating loop in B. Two Loops I/loop I/loop in two I/loop in each 110 60 (Below 50% of Full Power) operating loops operating loop C31 be 0 G

rcn acn O

O)

TABLE 3.5.1-1 (Continued)

TABLE NOTATION With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal.

A "TRIP" will stop all rod withdrawal.

Startup rate circuit enabled at 10-4% reactor power.

The provisions of Specification 3.0.4 are not applicable.

Below the Source Range High Voltage Cutoff Setpoint.

1. 1. 1. Below the P-7 (At Power Reactor Trip Defeat) Setpoint.

Above the P-7 (At Power Reactor Trip Defeat) Setpoint.

1. 1. 1. 1. 1. Above the P-8 Setpoint.

ACTION STATEMENTS ACTION 1 -

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 be in 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 />.

ACTION 2 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are met:

a. The inoperable channel is placed 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 />.

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be returned to the untripped condition for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of other channels per Specification 4.1.

ACTION 3 -

With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the Source Range High Voltage Cutoff Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the Source Range High Voltage Cutoff Setpoint.

b. Above the Source Range High Voltage Cutoff Setpoint but below 10 percent of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10 percent of RATED THERMAL POWER.

However, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.1, provided the other channel is OPERABLE.

ACTION 4 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.

SAN ONOFRE - UNIT 1 3.5-4 AMENDMENT NO:

55, 58, 83, 117, 118, 121, 128, 130

ACTION 5 -

With the number of OPERABLE channels one less than the Minimm Channels OPERABLE requirement, verify compliance with the SHUTDCHN MARGIN requirements of Specification 3.5.2 as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

ACTION 6 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required OPERATIONAL TEST provided the inoperable channel. is placed in the tripped condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

ACTION 7 -

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.

ACTION 28 -

With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirements, within one hour reduce THERMAL POWER such that Tave is less than or equal to 551.S'F, and place the rod control system in manual mode.

ACTION 29 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirements, 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 />; however, one channel may be removed from servtce-for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.1, provided the other channel is OPERABLE.

SAN ONOFRE - UNIT 1 3.5-5 AMENDMENT NO:

83.

117, 130

4.1.1 OPERATIGAL SAFETY ITEMS

~APLTCILIT:

Applies to surveillance requirements for items directly related to Safety Standards and Limiting Conditions for Operation.

To specify the minimum frequency and type of surveillance to be applied to plant equipment and conditions.

SPECIFICATION:

A. Reactor Trip System instrumentation shall be checked, tested, and calibrated as indicated in Table 4.1.1.

B. Equipment and sampling tests shall be as specified in Table 4.1.2.

C. The specific activity and boron concentration of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.1.2., Item la.

0. The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.1.2.,

Item lb.

E. All control rods shall be determined to be above the rod insertion limits shown in Figure 3.5.2.1 by verifying that each analog detector indicates at least 21 steps above the rod insertion limits, to account for the instrument inaccuracies, at least once per shift during Startup conditions with Keff equal to or greater than one.

F. The position of each rod shall be determined to be within the group demand limit and each rod position indicator shall be determined to be OPERABLE by verifying that the rod position indication system (Analog Detection System) and the step counter indication system (Digital Detection System) agree within 35 steps at least once per shift during Startup and Power Operation except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the rod position indication system (Analog Detection System) and the step counter indication system (Digital Detection System) at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

G. During MODE I or 2 operation each rod not fully inserted in the core shall be determined to be OPERASLE by movement of at least 10 steps in any one direction at least once per 31 days.

H. Instrumentation shall be checked, tested, and calibrated as indicated in Table 4.1.3.

SAN ONOFRE - UNIT 1 4.1-1 AMENDMENT NO:

29, 56, 83.

117, 130.

TABLE 4.1.1 (A

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEIlLANCE i1QUIREMENTS o

TRIP ACTUATING DEVICE CHANNEL CHANNEL CHANNEL OPERATIONAL ACTUATION FUNCTIONAL UNIT

.lCK.

CALIBRATION TEST1_

TEST LOGIC TEST

1. Manual Reactor Trip N.A.

N.A.

N.A. L N.A

2.

Power Range, Neutron Flux S

0 (2.3)

N N.A.

N.A.

R (3.4)

3.

Power Range. Neutron Flux.

N.A.

N.A.

N N.A.

N.A.

Dropped Rod Rod Stop

4.

Intermediate Range.

S R (3.4)

S/U (1).

N.A.

N.A Neutron Flux N

5.

Source Range, Neutron Flux S

R (3)

S/U (1).

N.A.

N.A.

N

6.

NIS Coincidentor Logic N.A.

N.A.

N.A.

N.A.

N (5)

1.

Pressurizer Variable Low S

R I N.A.

N.A.

Pressure

8.

Pressurizer Pressure S

R N

N.A.

N.A.

9.

Pressurizer Level 5

R N

N.A.

N.A.

i'

10.

Reactor Coolant Flow S

Q N.A.

N.A.

I1. Steam/Feedwater Flow S

R N

N.A.

N.A.

Z Mismatch

-4 o

12.

Turbine Trip-Low Fluid N.A.

N.A.

.A.

S/U (1.6)

N.A.

Oil Pressure

13.

Reactor Coolant Pump Breaker S

R R

N.A.

N.A.

P)

Position*

• Applicable to Item 6 In Table 2.1 W

TABLE 4.1.1 (Continued)

TABLLAQITATION (1) -

If not performed in previous 31 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(4) -

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(5) -

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(6) -

Setpoint verification is not applicable.

SAN ONOFRE -

UNIT 1 4.1-3 AMENDMENT NO:

117, 130..