ML20086K510
| ML20086K510 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/17/1995 |
| From: | Rosenblum R SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20086K501 | List: |
| References | |
| NUDOCS 9507200176 | |
| Download: ML20086K510 (18) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDIS0N COMPANY, ET AL. for a Class 103 License to DOCKET N0. 50-361 Acquire, Possess, and Use a Utilization Facility as Part of Unit No. 2 of the Amendment Application No. 139 j
San Onofre Nuclear Generating Station 1
SOUTHERN CALIFORNIA EDIS0N COMPANY, ET AL, pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 139.
i This amendment application consists of Proposed Change Number (PCN) NPF-10-434 to Facility Operating License No. NPF-10.
PCN llPF-10-434 is a request to revise San Onofre Unit 2 Technical Specification (TS) 3/4.3.1, " Reactor Protective Instrumentation," and TS 3/4.3.2, " Engineered Safety Feature Actuation System Instrumentation," and their associated Bases. This request will revise certain Plant Protective System instruaentation surveillance intervals to 120 days.
950720o176 950717 PDR ADOCK 05000361 P
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day of daL[
, 1995.
Subscribed on this /7M r
Respectfully submitted, SOUTHERN CALIFORNIA EDIS0N COMPANY By:
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Richard M. Rosenblum Vice President
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State of California of Orange fy g v County //? / 75 before me,BA sa4 A.heedur/V. * #, personally On
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appeared ///cwo M. Rescuamm, personally known to me to be the person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of which the person acted, I
executed the instrument.
WITNESS my hand and official seal.
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li U-UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDIS0N COMPANY, ET AL. for a Class 103 License to DOCKET NO. 50-362 Acquire, Possess, and Use a Utilization Facility as Part of Unit No. 3 of the Amendment Application No. 123 San Onofre Nuclear Generating Station SOUTHERN CALIFORNIA EDIS0N COMPANY, ET AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 123.
This amendment application consists of Proposed Change Number (PCN) NPF-15-434 to Facility Operating License No. NPF-15.
PCN NPF-15-434 is a request to revise San Onofre Unit 3 Technical Specification (TS) 3/4.3.1, " Reactor Protective Instrumentation," and TS 3/4.3.2, " Engineered Safety Feature Actuation System Instrumentation," and their associated Bases.
This request
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will revise certain Plant Protective System instrumentation surveillance intervals to 120 days.
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Subscribed on this /?rh day of EuCV 1995.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY l
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instrument the person, or the entity upon behalf of which the person acted, executed the instrument.
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ENCLOSURE I DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-10/15-434 P
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DESCRIPTION AND SAFETY ANALYSIS j
OF PROPOSED CHANGE NPF-10/15-434 This is a request to revise Surveillance Requirement (SR) 4.3.1.1 and SR 4.3.1.2 of Technical Specification (TS) 3/4.3.1, " Reactor Protective Instrumentation," and SR 4.3.2.1 of TS 3/4.3.2, " Engineered Safety Feature Actuation System Instrumentation," for San Onofre Units 2 and 3.
The Bases of TS 3/4.3.1 and 3/4.3.2, " Reactor Protective and Engineered Safety Features Actuation System Instrumentation," will also be revised.
Current Units 2 and 3 Technical Specifications Existing Specifications:
Unit 2: See Attachment "A" Unit 3: See Attachment "B" Proposed Specifications:
Unit 2: See Attachment "C" Unit 3: See Attachment "D" Revised Technical Specification Improvement Program (PCN-299) Technical Specifications l
Proposed Revision to PCN-299, Supplements 1 though 4:
Unit 2: See Attachment "E" Unit 3: See Attachment "F"
1.0 DESCRIPTION
OF CHANGES:
This amendment request is to increase the surveillance interval from 92 days sequential to 120 days staggered for certain Plant Protective j
System (PPS) instrumentation in Surveillance Requirement (SR) 4.3.1.1 and SR 4.3.1.2 of Technical Specification (TS) 3/4.3.1, " Reactor Protective Instrumentation," and SR 4.3.2.1 of TS 3/4.3.2, " Engineered I
Safety Feature Actuation System Instrumentation." SRs 4.3.1.1, 4.3.1.2, and 4.3.2.1 require in part that certain Plant Protective System instrumentation be demonstrated operable by the performance of a channel functional test at least once each 92 days.
This amendment request is to revise this interval to a 120-day staggered test surveillance interval.
A new notation "0" and footnote "0" will be added to TS tables 4.3-1 and 4.3-2 to denote the new test frequency of 120 days staggered. This notation would require a test frequency of 120 days on a staggered test basis (i.e., one Reactor Protective System (RPS)/ Engineered Safety Feature Actuation System (ESFAS) channel would be tested every 30 days, rotating through all four channels every 120 days).
" Staggered test basis" is a defined term in the Technical Specifications.
SR 4.3.1.1 requires the linear power subchannel gain amplifiers be calibrated monthly and the PPS bistables be calibrated once each 92
days.
This amendment request is to revise the calibration interval for the linear power subchannel gain amplifiers and associated nuclear instrumentation to once each 120 days sequential and the PPS bistables to a 120-day staggered test basis. SR 4.3.1.2 requires the logic for the bypasses be demonstrated operable prior to each reactor startup unless performed during the preceding 92 days. This amendment request is to revise the 92 day requirement to 120 days.
SRs 4.3.1.1 and 4.3.2.1 also require the reactor protective system and ESFAS logic be demonstrated operable by the performance of a channel functional test at least once each 92 days. This amendment request will revise this interval to 120 days. A new notation "**" and footnote "**"
will be added to TS Tables 4.3-1 and 4.3-2 to denote the new test frequency of once each 120 days.
The Bases of TS 3/4.3.1 and 3/4.3.2, " Reactor Protective and Engineered Safety Features Actuation System Instrumentation," will also be revised to reflect the new surveillance intervals.
2.0 BACKGROUND
1 2.1 Technical Specifications j
SRs 4.3.1.1, 4.3.1.2, and 4.3.2.1 require in part that each instrumentation channel be demonstrated operable by the performance of a channel functional test (CFT) in accordance with the surveillance interval frequency specified in Table 4.3-1 and Table 4.3-2.
The SRs 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 specified frequencies are sufficient to demonstrate this capability.
2.2 P.PS and ESFAS Instrumentation Systems The RPS Instrumentation and ESFAS Instrumentation together form the PPS.
The PPS is required to provide for the protection and mitigation of accident and transient conditions.
The RPS instrumentation consists of transmitters, calculators, logic, and other equipment necessary to monitor selected Nuclear Steam Supply System conditions and to effect reliable and rapid reactor shutdown (reactor trip) when monitored conditions approach specified safety system limits.
Four measurement channels with physical and electrical separation are provided for each parameter used in the direct generation of trip signals, with the exception of Control Element Assembly (CEA) position. A two out of four coincidence of like trip signals is required to generate a reactor trip signal.
The fourth channel is provided as a spare and allows bypassing of one channel while maintaining a two out of three 2
system. The reactor trip signal de-energizes the CEA drive mechanism coils, allowing all CEAs to drop into the core.
The ESFAS is designed to actuate engineered safeguards equipment intended to limit equipment damage, to provide protection for Control Room personnel, to minimize off-site radiation releases, and to mitigate the consequences of postulated accidents.
2.3 Channel Functional Testina The Channel Functional Test (CFT) verifies the alarm and/or trip actuation points for the channel as specified by the TSs. The CFT requires that a simulated test signal be applied and that the trip functions occur at the proper value of the simulated signal.
For the PPS this requirement is implemented by using installed test equipment to simulate a signal at the input to the bistable trip unit. The input is varied until the bistable actuates.
The value of the signal at the actuation point is then recorded and compared to the test acceptance criteria.
2.4 Asea Brown Boveri - Combustion Engineerina (ABB-CE) RPS/ESFAS Extended Test Interval Evaluation The current quarterly surveillance interval for functional units of the PPS is based on an analysis presented in Combustion Engineering Owners Group (CE0G) Topical Report, CEN-327-A, "RPS/ESFAS Extended Test Interval Evaluation," and Supplement I thereto. This topical report, including Supplement 1, was approved for use by member CE0G utilities by the NRC Safety Evaluation Report (SER) in the November 6, 1989 letter to the CEOG Chairman, Mr. E. Sterling.
The NRC found the CEN-327-A topical report provided an acceptable generic basis to support plant specific TS changes for extending both RPS and ESFAS CFT intervals from monthly to quarterly.
Probabilistic Risk Assessment (PRA) techniques were utilized to demonstrate that the proposed surveillance interval extensions did not result in increased plant risk when compared with current TS requirements. As also stated in the SER, the effects of plant instrument drift in both the sensors and instrument strings were not addressed by the topical report, and would be required for an individual plant submittal.
2.5 RPS/ESFAS Functional Units TS 3/4.3.1, Table 4.3-1, lists the following functional units for RPS instrumentation:
e 1.
Linear Power Level--High 3.
Logarithmic Power Level--High 4.
Pressurizer Pressure--High 3
5.
Pressurizer Pressure--Low 6.
' Containment Pressure--High 7.
Steam Generator Pressure--Low 8.
Steam Generator Level--Low 9.
Local Power Density--High l
- 10. Departure from Nucleate Boiling Ratio (DNBR)--Low
- 11. Steam Generator Level--High
- 12. Reactor Protection System Logic
- 13. Reactor Trip Breakers
- 14. Core Protection Calculators (CPCs)
- 15. Control Element Assembly Calculators (CEACs) 16.
Reactor Coolant Flow--Low
- 17. Seismic--High 18.
Loss of Load TS 3/4.3.2, Table 4.3-2, lists the following functional units for ESFAS instrumentation:
1.
Safety Injection Actuation Signal (SIAS) 2.
Containment Spray Actuation Signal (CSAS)
Main Steam Isolation Signal (MSIS)gnal (CIAS)
Containment Isolation Actuation Si 3.
4.
5.
Recirculation Actuation Signal (RAS) 6.
Containment Cooling Actuation Signal (CCAS) 7.
LossofPower(LossofVoltageSignal(LOVS), Sustained J
Degraded Voltage Signal (SDVS), or Degraded Grid Voltage with SIAS Signal (DGVSS))
8.
Emergency Feedwater Actuation Signal (EFAS) 9.
Control Room Isolation Signal (CRIS)
- 10. Toxic Gas Isolation Signal (TGIS) 11.
Fuel Handling Isolation Signal (FHIS)
- 12. Containment Purge Isolation Signal (CPIS)
RPS/ESFAS Surveillances Affected By This Amendment:
In addition to the functional units specifically addressed in CEN-327-A (which are RPS items 2 through 12 and 14 through 18 and ESFAS items 1 through 5 and 8), the CCAS and Seismic--High functional units also will have their surveillance intervals i
revised by approval of this amendment request from 92-day sequential to 120-day staggered.
The Nuclear Instrumentation (NI) i linear subchannel gain amplifiers and associated nuclear instrumentation will have their surveillance intervals revised by approval of this amendment request from 30-day sequential to 120-day sequential. The PPS bistables will have their surveillance intervals revised by approval of this amendment request from 92-day sequential to 120-day staggered.
The surveillance intervals for each of these additional functional units are being extended because they are needed to capture the benefits of the surveillance interval extension for the functional units addressed in CEN-327-A (i.e., reduction in reactor trip hazard and increase in PPS system avail. ability).
4
l RPS/ESFAS Surveillances Not Affected By This Amendment:
This amendment request does not affect Loss of Power, CRIS, TGIS, FHIS, or CPIS functional unit CFT surveillance intervals. This amendment request also does not affect the CFT surveillance interval for the Manual Reactor Trip or Reactor Trip Breakers j
functional units.
Twenty-four Month Calibration Interval Effects:
The NRC Safety Evaluation Report for Amendments 88 and 78, for Units 2 and 3, respectively, requires that future San Onofre submittals which request an extension of the RPS and ESFAS monthly channel functional test confirm that the bases for the 24-month calibration interval approved in Amendment Nos. 88 and 78 is not compromised.
This amendment request does not change the margin or error allowances related to the channel calibration interval of 24 i
months. Therefore, this amendment request does not compromise the bases for the 24-month calibration interval.
3.0 DISCUSSION
This amendment request is to extend the CFT surveillance interval for certain RPS and ESFAS instrumentation channels.
Increasing the CFT surveillance interval for these RPS and ESFAS instrumentation channels has four principle benefits: 1) RPS and ESFAS system availability will be increased (Tables 4.1-1 and 4.2-1 of Enclosure 3 and Table S-1 of i provide a quantitative value for the increase in availability), 2) the potential for inadvertent ESFAS actuation and reactor trip from performing this surveillance testing will be decreased, 3) burdens on plant personnel involved in performing CFTs will be reduced significantly, and 4) there will be a projected total cost savings of over $1,000,000 over the life of the two units.
4 The current quarterly surveillance interval for functional units of the PPS is based on an analysis presented in CE0G Topical Report CEN-327-A including Supplement 1.
This report was approved for use by member CE0G utilities by the NRC Safety Evaluation Report (SER) on November 6, 1989.
In June 1989, based on preliminary feedback from the NRC with regard to CEN-327-A and its supplement, the CE0G Reliability and Availability Subcommittee requested that ABB-CE evaluate the impact of extending the RPS and ESFAS CFT intervals from 30-day sequential testing to 120-day staggered testing.
The results of these analyses, which have not been transmitted to the NRC for review and approval, were documented in draft i
report CE NFSD-576.
Based on the results of draft report CE NPSD-576, Edison asked ABB-CE to revise the draft report to provide a formal document specifically for San Onofre Units 2 and 3.
This document, "RPS/ESFAS Extended Test Interval Evaluation," (S023-944C-29-0) evaluated the impact for San Onofre Units 2 and 3 of extending the RPS and ESFAS CFT intervals from 5
30-day sequential testing to 120-day staggered testing.
It also evaluates the impact of extending the RPS and ESFAS CFT intervals from 90-day sequential testing to 120-day staggered testing. The analysis demonstrates that extending the CFT surveillance intervals from 30 days to 120 days results in a change in unavailability which is the same or smaller than the change associated with extending the test interval from 30 days to 90 days.
To fully realize the benefits of the RPS/ESFAS CFT interval extension the Nuclear Instrumentation excore detectors linear subchannel gain and associated nuclear instrumentation monthly verifications must also be extended to 120 days. Since these verifications were not included in the scope of CEN-327-A or the Units 2 and 3 specific report Edison commissioned a plant specific nuclear instrumentation system unavailability study to support extension of the monthly verifications.
The system unavailability study was performed by Pickard, Lowe and Garrick, Inc. and was issued as Report PLG-0575 (5023-941-91),
" Methodology for Developing Risk-Based Surveillance Programs for Safety-Related Equipment at San Onofre Nuclear Generating Station Units 2 and 3."
This report shows the surveillance intervals for the nuclear instrumentation linear power subchannel gain amplifiers and associated nuclear instrumentation can be extended to a 120-day sequential interval with an increase in system availability.
The NRC Safety Evaluation Report for CEN-327-A required that each licensee confirm they have reviewed drift information for each channel involved in a test surveillance interval extension. Edison performed an evaluation of the effects of instrument drift on the RPS/ESFAS CFT and nuclear instrumentation calibration interval extension. The results of these evaluations found the drift values to be largely within allowable limits.
For those cases where the values were outside allowable limits appropriate actions are being taken to accommodate the new values.
A detailed discussion of each of the above evaluations is provided in the following sections.
3.1 ABB-CE San Onofre Units 2 and 3 Specific RPS/ESFAS Extended Test J_nterval Evaluation ABB-CE performed a San Onofre Units 2 and 3 specific evaluation for extending the RPS/ESFAS CFT surveillance interval from 92 days sequential to 120 days staggered. This evaluation was derived from the RPS and ESFAS fault tree models developed for and presented in CEN-327-A.
Four RPS fault tree models were constructed, one for each of the ABB-CE RPS designs.
These models were constructed such that the components associated with the trip parameters were developed as separate subtrees. Thus, it was possible to analyze the RPS design utilized at San Onofre Units 2 and 3 on a trip parameter basis simply by constructing the appropriate subtrees to the main RPS fault tree presented in Appendix A of CEN-327-A.
6
Similarly, fault tree models were constructed for the appropriate ESFAS signals.
The same methodology used in CEN-327-A and its supplement for the ESFAS design is employed in an analysis ABB-CE performed for Edison.
The resulting RPS and ESFAS fault trees were evaluated using NUREG/CR-5111, " Integrated Reliability and Risk Analysis System (IRRAS) User's Guide - Version 2.0."
For CEN-327-A, the RPS and ESFAS fault trees were evaluated using the SETS code for cutset generation, and the KITT code for quantification. The original SETS RPS and ESFAS fault tree model input decks were loaded into IRRAS via its SETS interface.
These models were then evaluated and quantified using component failure rates based on a 30-day test interval. The results of these quantifications were compared to the equivalent results contained in CEN-327-A to verify that they were the same.
Next, the failure rates for the bistables, the bistable relays, the logic matrix relays, and the trip path relays were changed to reflect a 120-day staggered test interval and the fault trees were requantified to determine the system reliability for a 120-day test interval with staggered testing.
The analysis also verified the acceptability of a 120-day sequential CFT interval for the matrix and trip path relay portion of the system.
(For clarification, both the solid state and K relays are the trip path relays for the RPS and the initiation relays are regarded as the trip path relays for the ESFAS.)
The Seismic--High functional unit was not explicitly addressed by CEN-327-A. However, ABB-CE has determined that the similarity in j
the design of the seismic trip function bistable to the Loss of i
Load trip function bistable would allow the use of the same analysis to justify extending the surveillance interval.
CEN-327-A did not specifically address CCAS.
CCAS, however, is included in this amendment request. The CCAS and SIAS share the same type of bistables and are designed similarly. Therefore, the CCAS can be encompassed in the SIAS analysis.
This has been verified by ABB-CE.
The fault tree analysis results are presented as system unavailabilities.
(Unavailability is a measure of system reliability--a decrease in unavailability results in an increase in reliability.) The RPS unavailability is the probability that the RPS is unavailable to perform its function of tripping the reactor when required. The ESFAS unavailability is the probability that the ESFAS fails to actuate specific ESFAS components.
The analysis demonstrates that extending the CFT surveillance interval for the RPS and ESFAS bistables from 92-day sequential to 120-day staggered testing and extending the CFT surveillance interval for the matrix and trip path relays from 92-day sequential to 120-day sequential testing results in a change in 7
unavailability which is the same or smaller than the change associated with extending the test interval from 30 days sequential to 90 days sequential. The change in unavailability for 120-day sequential and staggered testing meets the acceptance criteria in that the NRC has already accepted the change in unavailability for 90-day sequential testing.
The ABB-CE San Onofre Units 2 and 3 specific report "RPS/ESFAS Extended Test Interval Evaluation for 120 Days Staggered Testing" is provided as.
3.2 Pickard. Lowe and Garrick. Inc.. San Onofre Units 2 and 3 Specific Nuclear Instrumentation Linear Subchannel Gain Extended Test Interval Evaluation Table 4.3-1 of the current Units 2 and 3 Technical Specifications requires monthly verification that the Nuclear Instrumentation (NI) linear subchannel gains for the excore detectors are consistent with the values used to establish the shape annealing matrix element in the Core Protection Calculators.
Because the NI channel drawers, which include the linear subchannel gains, were not included in CEN-327-A, Edison commissioned a plant specific i
nuclear instrumentation system unavailability study to support extension of the channel calibration interval for the nuclear instrumentation including the linear subchannel gains settings.
The system unavailability study was performed by Pickard, Lowe and Garrick, Inc. (PLG) and was issued as Report PLG-0575 (S023-941-91), " Methodology for Developing Risk-Based Surveillance Programs for Safety-Related Equipment at San Onofre Nuclear Generating Station Units 2 and 3."
(This report is provided as Enclosure 4.)
In this study, the unavailability of the excore nuclear instrumentation safety channel drawers to produce a proper voltage output when required for reactor trip was evaluated as the measure of risk. A quantitative evaluation of the surveillance test interval was accomplished by estimating, from both generic and plant specific data, the time-dependent and test-related failure parameters. The SOCRATES computer code was then used to conduct time-dependent unavailability analyses to determine the sensitivity of average system unavailability to surveillance test interval, channel bypass time, and between test time-related
" standby" failures.
The results of the quantitative evaluation using the best estimate values of the system parameters is provided in the PLG-0575 report. This report indicates that the surveillance intervals for the nuclear instrumentation linear power subchannel gain amplifiers and associated nuclear instrumentation can be extended to a 120-day sequential interval with no increase in system unavailability.
The quantitative results actually document an increase in system availability with the surveillance interval extension.
8
3.3 Southern California Edison Instrument Drift Analyses The NRC Safety Evaluation Report for CEN-327-A required that each licensee confirm they have reviewed drift information for each channel involved in a test surveillance interval extension and have determined that drift occurring in that channel over the period of the extension will not cause the setpoint value to exceed the allowable value as calculated for that channel by their setpoint methodology. Because this requirement must be met for extension of the RPS/ESFAS CFT intervals Edison performed an evaluation of the effects of instrument drift on the RPS/ESFAS CFT interval extension.
In addition, a drift calculation has been performed for the nuclear instrumentation channels, including the linear subchannel gain amplifiers.
The drift calculation for the nuclear instrumentation linear subchannel gains and associated nuclear instrumentation provides the results for analysis of the amplifier drift over a 120-day period, based on plant specific data from San Onofre Units 2 and 3.
l The methodology for each of these calculations is similar.
Edison began by identifying the components which are subject to drift and are affected by an increased CFT surveillance interval.
These components are then assessed to determine the suitability of extending the CFT surveillance interval based on the historically experienced drift on a "95/95" industry standard basis.
Bounding values for drift were developed, and the probability that the trip setpoint would remain within these bounding values was calculated.
This resulted in a percentage of tests where the trip setpoint 1
would fall within the bounding values.
To ensure-the probability was conservative, minimum values of the probability were calculated so that there would be at least a 95% confidence that the true probability was greater than the minimum value.
For the RPS/ESFAS bistables (Edison Calculation No. J-SBA-001),
the results for all cases were within the drift values assumed and allowed in the RPS/ESFAS and CPC uncertainty calculations, which include the subject bistables. The drift calculation performed for the linear subchannel gains settings (Edison Calculation No.
J-SEA-012) resulted in drift values for 120 days which were larger than the allowances in the RPS/ESFAS and CPC instrument uncertainty and setpoint calculations for Units 2 and 3.
The RPS/ESFAS and CPC instrument uncertainty calculations (S023-944-C50-0 and S023-944-C90-0, respectively) have been revised to include the larger allowances. The RPS/ESFAS instrument setpoint calculations and CPC uncertainty analysis has been updated.
The longer 120-day calibration frequency was included in conjunction with other updates. These updates resulted in a net increase in the CPC overpower operating margin. The Edison drift calculations are provided as Enclosure 5.
The drift calculation for the Reed Switch Position Transmitter 15 volt power supplies identified that one of four power supplies 9
would not be within the drift values assumed and allowed in the CPC uncertainty calculations for a 120-day CFT interval.
(The drift value was acceptable for the current 90-day CFT interval.)
This power supply has been replaced.
4.0 DISCUSSION OF CHANGES TO PCN-299:
PCN-299 implements the Technical Specification Improvement Project which incorporates the recommendations of NUREG-1432, " Standard Technical Specifications Combustion Engineering Plants." PCN-299 was submitted to the NRC for review on December 30, 1993.
PCN-299 changes the verification that the Nuclear Instrumentation (NI) linear subchannel gains for the excore detectors are consistent with the values used to establish the sha)e annealing matrix element in the Core Protection Calculators from mont11y to 92 days. This amendment request will extend this verification requirement from 92 days to 120 days.
This extension to 120 days will yield a reduction in system unavailability over the existing monthly verification requirement.
The system unavailability on a 120-day verification interval, although larger (i.e., an increase of about IE-6 per demand), is not significantly different from the system univailability on a 92-day verification interval.
PCN-299 has a different definition for :>TAGGERED TEST BASIS than the current Technical Specifications (prior to PCN-299 implementation). The current Technical Specification definition of a 120-day STAGGERED TEST BASIS divides the 120-day test interval into four intervals of thirty days (one interval per train) with testing of each train in turn on a 30-day interval until each train has been tested.
This definition is equivalent to the PCN-299 definition of a 30-day STAGGERED TEST BASIS, that is each of the four trains will be tested in turn on a 30-day interval until all four trains are tested, i
The changes identified in Attachments E and F reflect changes required i
to implement the PPS CFT interval extension to a 120-day STAGGERED TEST BASIS as defined in the current Technical Specifications (prior to PCN-299 implementation). Therefore, the PCN-299 changes identify the CFT intervals as 30-days on a STAGGERED TEST BASIS.
(Four channels over 120 days.)
The CEACs have only two channels, thus yielding CFT intervals of 60-days on a STAGGERED TEST BASIS.
(Two channels over 120 days.)
The PCN-299 definition for sequential test basis remains unchanged from the current Technical Specification definition.
10
4 t
5.0' SAFETY' ANALYSIS:
~
The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any one of the following areas:
1.
Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or.
consequences of any accident previously evaluated?
Response: No y
The proposed change would extend the current sequential Channel-Functional Test (CFT) surveillance interval for Plant Protective System (PPS) instrumentation and Nuclear Instrumentation (NI).
This change does not involve any changes to plant equipment or operation. The proposed change actually maintains or decreases the PPS system unavailability.
PPS uncertainty and setpoint modifications will account for the new surveillance interval.
Therefore, the proposed change will not involve a significant increase in the probability or consequences of any 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 j
accident from any previously evaluated?
Response: No This amendment request does not involve any change to plant equipment or operation. The PPS system is used for monitoring and mitigation of evaluated accidents.
Increasing the availability of the PPS system, as proposed in this amendment request, will not create the possibility of a new or different kind of accident 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 This amendment does not change the manner in which safety limits, limiting safety settings, or limiting conditions for operation are determined. This amendment request will increase Reactor Protective System and Engineered Safety Features Actuation System availability.
Therefore, this amendment will not involve a significant reduction in a margin of safety.
6.0 SAFETY AND SIGNIFICANT HAZARDS DETERMINATION:
Based on the above Safety Analysis, it is concluded that: (1) the I
proposed change does not constitute a significant hazards consideration 11 I
f 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. Moreover, because this action does not involve a significant hazards consideration, it will also not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.
I 12 i
ATTACHMENT A EXISTING TECHNICAL SPECIFICATIONS AND BASES UNIT 2 I-F
_