LR-N21-0006, Application to Revise Technical Specifications to Adopt TSTF-569 Revision of Response Time Testing Definitions

Application to Revise Technical Specifications to Adopt TSTF-569 Revision of Response Time Testing Definitions

ML21047A238
Person / Time
Site:	Salem
Issue date:	02/16/2021
From:	Sharbaugh D Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR S19-06, LR-N21-0006
Download: ML21047A238 (16)

Text

10 CFR 50.90 LR-N21-0006 LAR S19-06 February 16, 2021 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

Application to Revise Technical Specifications to Adopt TSTF-569 Revision of Response Time Testing Definitions In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Salem Generating Station (Salem) Units 1 and 2.

PSEG requests adoption of TSTF-569, "Revise Response Time Testing Definition," into the Salem Units 1 and 2 TS. The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time.

The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked up to show the proposed changes. Attachment 2 provides existing TS Bases pages marked to show the proposed changes for information only.

PSEG requests approval of this license amendment request (LAR) in accordance with standard NRC approval process and schedule. Once approved, the amendment will be implemented within 60 days from the date of issuance.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

If you have any questions or require additional information, please contact Mr. Michael Wiwel at 856-339-7907.

LR-N21-0006 10 CFR 50.90 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on __),--1/_1Iftv2----+-(

(Date)

Respectfully,

])J David Sharbaugh Site Vice President Salem Generating Station

Enclosure:

Description and Assessment Mark-up of Proposed Technical Specification Pages Mark-Up of Proposed Technical Specification Bases Pages - For Information Only cc: Administrator, Region I, NRC NRC Project Manager NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Station Compliance Commitment Tracking Coordinator

LR-N21-0006 LAR S19-06 Enclosure Description and Assessment Table of Contents

1.0 DESCRIPTION

.................................................................................................................. 1 2.0 ASSESSMENT .................................................................................................................. 1 2.1 Applicability of Safety Evaluation ................................................................................ 1 2.2 Variations .................................................................................................................... 1

3.0 REGULATORY ANALYSIS

............................................................................................... 2 3.1 No Significant Hazards Consideration Determination ................................................. 2 3.2 Conclusions ................................................................................................................. 3 4.0 ENVIONMENTAL EVALUATION ...................................................................................... 3 ATTACHMENTS:

1. Mark-up of Proposed Technical Specification Pages

2. Mark-up of Proposed Technical Specification Bases Pages - For Information Only

LR-N21-0006 LAR S19-06 Enclosure

1.0 DESCRIPTION

PSEG Nuclear LLC (PSEG) requests adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Salem Units 1 and 2 Technical Specifications (TS). The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation PSEG has reviewed the safety evaluation for TSTF-569 provided to the Technical Specifications Task Force (TSTF) in a letter dated August 14, 2019. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-569. As described herein, PSEG has concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to Salem Units 1 and 2 and justify this amendment for the incorporation of the changes to the Salem TS.

2.2 Variations The current Salem TS DEFINITIONS for Engineered Safety Function (ESF) Response Time and Reactor Trip System (RTS) Response Time do not encompass the full text found in the Standard Improved Technical Specifications in NUREG-1431 for these DEFINITIONS. The text in the NUREG-1431 DEFINITIONS pertaining to verification of response times for selected components and measurement by means of any series of sequential or overlapping steps is also being added to the Salem TS definition in addition to the change associated with implementing TSTF-569.

Inclusion of the additional NUREG-1431 DEFINITION text for the ESF and RTS RESPONSE TIME aligns the Salem TS DEFINITION with the Salem Licensing Basis relative to the TS response time surveillance for RTS (SR 4.3.1.1.3) and ESF (SR 4.3.2.1.3). The allowance to demonstrate response time by any series of sequential, overlapping or total channel test measurements was provided in the initial TS Bases issued with Salem Facility Operating Licenses. The SRs, and the associated TS Bases, were changed per Amendments 260 and 241 for Salem Unit 1 and 2 respectively (Reference ML033010379) to allow verification of response time in lieu of demonstration through physical testing. The changes implemented by these TS Amendments conformed to the changes approved in WCAP-13632-P-A, Revision 2 which contains the technical basis and methodology for eliminating Response Time Testing requirements on selected pressure sensing instruments.

The Model Application contained in TSTF-569 contains the final, camera-ready, version of the revised TS change along with a markup of the affected TS. PSEG has elected to withhold this final version of the TS change to avoid any potential implementation conflicts.

1

LR-N21-0006 LAR S19-06 Enclosure

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination PSEG requests adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Salem Unit 1 and Unit 2 Technical Specifications (TS). The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time.

PSEG has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC-approved methodology and apply a bounding response time for some components in lieu of measurement. The requirement for the instrumentation to actuate within the response time assumed in the accident analysis is unaffected.

The response time associated with the RTS and ESF instrumentation is not an initiator of any accident. Therefore, the proposed change has no significant effect on the probability of any accident previously evaluated.

The affected RTS and ESF instrumentation are assumed to actuate their respective components within the required response time to mitigate accidents previously evaluated.

Revising the TS definition for RTS and ESF instrumentation response times to allow an NRC-approved methodology for verifying response time for some components does not alter the surveillance requirements that verify the RTS and ESF instrumentation response times are within the required limits. As such, the TS will continue to assure that the RTS and ESF instrumentation actuate their associated components within the specified response time to accomplish the required safety functions assumed in the accident analyses. Therefore, the assumptions used in any accidents previously evaluated are unchanged and there is no significant increase in the consequences.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC-approved methodology and apply a bounding response time for some components in lieu of measurement. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The proposed change does not alter any assumptions made 2

LR-N21-0006 LAR S19-06 Enclosure in the safety analyses. The proposed change does not alter the limiting conditions for operation for the RTS or ESF instrumentation, nor does it change the Surveillance Requirement to verify the RTS and ESF instrumentation response times are within the required limits. As such, the proposed change does not alter the operability requirements for the RTS and ESF instrumentation, and therefore, does not introduce any new failure modes.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC-approved methodology and apply a bounding response time for some components in lieu of measurement. The proposed change has no effect on the required RTS and ESF instrumentation response times or setpoints assumed in the safety analyses and the TS requirements to verify those response times and setpoints. The proposed change does not alter any Safety Limits or analytical limits in the safety analysis. The proposed change does not alter the TS operability requirements for the RTS and ESF instrumentation. The RTS and ESF instrumentation actuation of the required systems and components at the required setpoints and within the specified response times will continue to accomplish the design basis safety functions of the associated systems and components in the same manner as before. As such, the RTS and ESF instrumentation will continue to perform the required safety functions as assumed in the safety analyses for all previously evaluated accidents.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in 3

LR-N21-0006 LAR S19-06 Enclosure individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

4

LR-N21-0006 LAR S19-06 Attachment 1 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification Page DEFINITION 1.12 - ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 DEFINITION 1.26 - REACTOR TRIP SYSTEM RESPONSE TIME 1-6 The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification Page DEFINITION 1.12 - ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 DEFINITION 1.26 - REACTOR TRIP SYSTEM RESPONSE TIME 1-6

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 DEFINITIONS and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion 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 greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF 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.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or.shutdown banks are at a position which is within the interval of 222 to 230 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 1 1-3 Amendment No.292

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME RTS 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3 and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 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 1.30 Not Used of measurement, response time may be verified for selected components SOURCE CHECK provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been 1.31 SOURCE CHECK evaluated shall be in accordance the withassessment qualitative an NRC approved methodology.

of channel response when the channel sensor is exposed to either {a) an external source of increased radioactivity, or(b) an internal source o f radioactivity (keep-alive source), or (c) an equivalent electronic source check.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - UNIT 1 1-6 Amendment No. 280

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 DEFINITIONS reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and ganuna energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF 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.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN l.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 230 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive g aseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 2 1-3 Amendment No. 276

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME RTS 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

The response time may be measured by means of any series of sequential, SOLIDIFICATION overlapping, or total steps so that the entire response time is measured. In lieu of 1.30 Not Used measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously SOURCE CHECK reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to either (a) an external source of increased radioactivity, or(b) an internal source of radioactivity (keep alive source), or (c) an equivalent electronic source check .

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - UNIT 2 1-6 Amendment No. 263

LR-N21-0006 LAR S19-06 Attachment 2 Mark-up of Proposed Technical Specification Bases Pages For Information Only The following Technical Specifications Bases pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification Page 3/4.3 Instrumentation B 3/4 3-1b The following Technical Specifications Bases pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification Page 3/4.3 Instrumentation B 3/4 3-1b

BASES of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types, and other components that do not have plant-specific NRC approval to use alternate means of verification, must be demonstrated by test. The allocation for sensor response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where response time could be affected is replacing the sensing assembly of a transmitter. INSERT Channel testing in a bypassed condition shall be performed without lifting leads or jumpering bistables.

The CHANNEL CALIBRATION surveillance for the Power Range Neutron Flux Function instrumentation is modified by Note 17. Note 17 states that in MODES 1 and 2 the SSPS input relays are excluded from this Surveillance when the installed bypass test capability is used to perform this surveillance. When the installed bypass test capability is used, the channel is tested in bypass versus tripped condition. To preclude placing the channel in a tripped condition, the SSPS input relays are excluded from this surveillance. The exclusion of the SSPS input relays from this test is intended to reduce the potential for an inadvertent reactor trip during surveillance testing. Therefore, the exclusion of the SSPS input relays from the surveillance is only applicable in MODES 1 and 2. The SSPS input relays must be included in the CHANNEL CALIBRATION surveillance at least once every 18 months.

The CHANNEL FUNCTIONAL TEST surveillances for the Power Range Neutron Flux and Power Range Neutron Flux High Positive Rate Function Instrumentation are modified by Note

18. Note 18 states that the SSPS input relays are excluded from this surveillance when the installed bypass test capability is used to perform this surveillance. When the installed bypass test capability is used, the channel is tested in a bypassed versus tripped condition. To preclude placing the channel in a tripped condition, the SSPS input relays are excluded from this surveillance. The exclusion of the SSPS input relays from this test is intended to reduce the potential for an inadvertent reactor trip during surveillance testing. The SSPS input relays must be included in the CHANNEL CALIBRATION surveillance at least once every 18 months.

The Intermediate Range (IR) Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup. This trip Function provides redundant protection to the Power Range (PR) Neutron Flux

- Low Setting trip Function. In MODE 1 below the P-10 setpoint, and in MODE 2 above the P-6 setpoint, when there is a potential for an uncontrolled RCCA bank rod withdrawal accident during reactor startup, the IR Neutron Flux trip must be OPERABLE. Above the P-10 setpoint, the PR Neutron Flux - High Setpoint trip function provides core protection for a rod withdrawal accident. In MODE 2 below the P-6 setpoint, the Source Range (SR) Neutron Flux Trip provides core protection for reactivity accidents. In MODE 3, 4, or 5, the IR Neutron Flux trip does not have to be OPERABLE because the control rods must be fully inserted and only the shutdown rods may be withdrawn. The reactor cannot be started up in this condition.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Action times for one inoperable IR channel allow for a slow and controlled power adjustment above P-10 or below P-6 and take into account the redundant capability afforded by the OPERABLE IR channel and its low probability of failure during this period.

SALEM - UNIT 1 B 3/4 3-1b Amendment No. 335 (PSEG Issued)

INSTRUMENTATION BASES measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements 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 WCAP. Response time verification for other sensor types, and other components that do not have plant-specific NRC approval to use alternate means of verification, must be demonstrated by test.

The allocation for sensor response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where response time could be affected is replacing the sensing assembly of a transmitter. INSERT Channel testing in a bypassed condition shall be performed without lifting leads or jumpering bistables.

The CHANNEL CALIBRATION surveillance for the Power Range Neutron Flux Function instrumentation is modified by Note 17. Note 17 states that in MODES 1 and 2 the SSPS input relays are excluded from this surveillance when the installed bypass test capability is used to perform this surveillance. When the installed bypass test capability is used, the channel is tested in bypass versus tripped condition. To preclude placing the channel in a tripped condition, the SSPS input relays are excluded from this surveillance. The exclusion of the SSPS input relays from this test is intended to reduce the potential for an inadvertent reactor trip during surveillance testing. Therefore, the exclusion of the SSPS input relays from the surveillance is only applicable in MODES 1 and 2. The SSPS input relays must be included in the CHANNEL CALIBRATION surveillance at least once every 18 months.

The CHANNEL FUNCTIONAL TEST surveillances for the Power Range Neutron Flux and Power Range Neutron Flux High Positive Rate Function Instrumentation are modified by Note

18. Note 18 states that the SSPS input relays are excluded from this surveillance when the installed bypass test capability is used to perform this surveillance. When the installed bypass test capability is used, the channel is tested in a bypassed versus tripped condition. To preclude placing the channel in a tripped condition, the SSPS input relays are excluded from this surveillance. The exclusion of the SSPS input relays from this test is intended to reduce the potential for an inadvertent reactor trip during surveillance testing. The SSPS input relays must be included in the CHANNEL CALIBRATION surveillance at least once every 18 months.

The Intermediate Range (IR) Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup. This trip Function provides redundant protection to the Power Range (PR) Neutron Flux

- Low Setting trip Function. In MODE 1 below the P-10 setpoint, and in MODE 2 above the P-6 setpoint, when there is a potential for an uncontrolled RCCA bank rod withdrawal accident during reactor startup, the IR Neutron Flux trip must be OPERABLE. Above the P-10 setpoint, the PR Neutron Flux - High Setpoint trip function provides core protection for a rod withdrawal accident. In MODE 2 below the P-6 setpoint, the Source Range (SR) Neutron Flux Trip provides core protection for reactivity accidents. In MODE 3, 4, or 5, the IR Neutron Flux trip does not have to be OPERABLE because the control rods must be fully inserted and only the shutdown rods may be withdrawn. The reactor cannot be started up in this condition.

SALEM - UNIT 2 B 3/4 3-1b Amendment No. 316 (PSEG Issued)

LR-N21-0006 LAR S19-06 Bases Insert:

INSERT The response time may be verified for components that replace the components that were previously evaluated in Rev. 2 of WCAP-13632-P-A provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants Only) Response Time Testing.