RC-18-0020, License Amendment Request - LAR 17-04110 Technical Specification Change Request for the Revision of the Surveillance Frequency of the Turbine Trip Functional Unit

License Amendment Request - LAR 17-04110 Technical Specification Change Request for the Revision of the Surveillance Frequency of the Turbine Trip Functional Unit

ML18094A189
Person / Time
Site:	05000300
Issue date:	04/03/2018
From:	Lippard G SCANA Corp, South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 17-04110, RC-18-0020
Download: ML18094A189 (16)

Text

George A. Lippard Vice President, Nuclear Operations 803.345.4810 A SCANA COMPANY April 3, 201 8 RC-18-0020 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST - LAR (17-04110)

TECHNICAL SPECIFICATION CHANGE REQUEST FOR THE REVISION OF THE SURVEILLANCE FREQUENCY OF THE TURBINE TRIP FUNCTIONAL UNIT.

Pursuant to 10 CFR 50.90, South Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for South Carolina Public Service authority, hereby requests a license amendment to change Functional Units 17.A and 17.B of Technical Specification (TS) Table 4.3-1, "Reactor Trip System Instrumentation Surveillance Requirements". The Trip Actuating Device Operational Test (TADOT) column of this table will be revised to delete the "S/U" frequency and replace it with a reference to Table Notation (8) for prior to entering MODE 1 whenever the unit has been in MODE 3. The proposed surveillance frequency does not alter the current provision that takes credit for a surveillance performance within the previous 31 days. contains a description and assessment of the proposed changes. Attachment 2 contains the marked-up version of the affected TS page. Attachment 3 contains the reprinted version of the affected TS pages.

This amendment request was evaluated and found to have no significant hazards for consideration, as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

Approval of the proposed amendment is requested by 10/25/2018 to support RF-24. Once approved, the amendment shall be implemented within 7 days.

The VCSNS Plant Safety Review Committee and the Nuclear Safety Review Committee have reviewed and approved the proposed change. SCE&G is notifying the State of South Carolina of this LAR by transmitting a copy of this letter and enclosure to the designated State Official in accordance with 10 CFR 50.91(b).

There are no new commitments contained in this submittal.

V. C. Summer Nuclear Station

• P. 0. Box 88

• Jenkinsville, South Carolina

• 29065

• F (803) 941-9776

• www.sceg.com

Document Control Desk RC-18-0020 CR-17-04110 Page 2 of 2 If there are any questions or if additional information is needed, please contact Michael S.

Moore at (803) 345-4752 I declare under penalty of perjury that the foregoing is true and correct.

BAB/GAL/wk : Description and Assessment of the Proposed Changes : Existing TS Pages Marked to Show the Proposed Changes : Revised (Clean) TS Pages c:

J. E. Addison NRC Resident Inspector K. M. Sutton S. E. Jenkins P. Led better NSRC RTS (CR-17-04110)

File (813.20)

PRSF (RC-18-0020)

W. K. Kissam J. B. Archie J. H. Hamilton G. J. Lindamood W. M. Cherry C. Haney S. A. Williams

Document Control Desk RC-18-0020 CR-17-04110 Page 1 of 8 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT 1 Description and Assessment of the Proposed Changes

Subject:

This evaluation supports a request to amend South Carolina Electric & Gas Company (SCE&G), Technical Specifications (TS) to modify the surveillance frequency of the turbine trip functional unit.

1

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, SCE&G, acting for itself and as an agent for South Carolina Public Service Authority, hereby requests Nuclear Regulatory Commission (NRC) review and approval to amend Operating License NPF-12 for Virgil C. Summer Nuclear Station (VCSNS) Unit 1.

VCSNS is proposing to revise Functional Units 17.A and 17.B of TS Table 4.3-1, "Reactor Trip System Instrumentation Surveillance Requirements". The Trip Actuating Device Operational Test (TADOT) column of this table would be revised to delete the "S/U" frequency and replace it with a reference to Table Notation (8) which would state: Prior to entering MODE 1 whenever the unit has been in MODE 3. The proposed surveillance frequency does not alter the current provision that takes credit for a surveillance performance within the previous 31 days.

This change is desired because of the significant risk of causing a Safety Injection (SI) during the performance of this surveillance test prior to reactor startup. The steam demand required for this surveillance causes a cooldown of the Reactor Coolant System (RCS), and with very low decay heat, produces a large cooldown and depressurization. This configuration increases the potential of causing a SI due to the rapid cooldown and depressurization of the Main Steam and RCS.

Additionally, this change will align the surveillance requirements and the mode requirement for the Turbine Trip TADOT with the TS 3/4.3.1 Table 3.3-1 Rx Trip System Instrumentation channels and interlocks mode requirement.

Document Control Desk RC-18-0020 CR-17-04110 Page 2 of 8 2

DETAILED DESCRIPTION 2.1 System Design and Operation The proposed change affects the mode required to perform the surveillance test on the reactor trip system instrumentation that is initiated by the turbine trip functions whenever the unit has been in MODE 3. The reactor trip on a turbine trip is actuated by 2 out of 3 logic from the emergency trip fluid pressure signals or by all closed signals from the turbine steam stop valves.

A turbine trip initiates a reactor trip when above the P-9 interlock (a power level less than or equal to 50% of rated thermal reactor power).

The reactor trip on turbine trip is an anticipatory trip input signal to the reactor protection system.

The turbine provides anticipatory trips to the reactor protection system from contacts which change position when the turbine stop valves close or when the turbine emergency trip fluid pressure goes below its setpoint. This trip is anticipatory in that it is not assumed to occur in any of the analyzed accidents in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). The reactor protection system cabinets which receive the inputs from the anticipatory trip sensors are seismically qualified as discussed in Section 3.10 of the UFSAR. The anticipatory trips thus meet IEEE Standard 279-1971, including separation, redundancy, single failure, and testability.

The current revision of the TS requires this test be performed prior to MODE 2. If it is performed during secondary plant heat up, i.e. with the Main Steam Isolation Valves (MSIV) open prior to reactor startup, the only heat input to the secondary system is that from decay heat and the reactor coolant pumps. When the Main Turbine is reset prior to performing this test, the Main Stop Valves open causing a sudden demand on the Main Steam supply. With the limited amount of heat available prior to reactor startup, this sudden demand will drop the Main Steam pressure with a corresponding drop in the Reactor Coolant System pressure. An SI will occur if the pressurizer pressure drops to 1850 PSIG or the steam line pressure drops to 675 PSIG.

Removing the startup requirement will allow the test to be performed with reactor power sufficient (up to 5%) to prevent a SI due to the rapid cooldown and depressurization of the Main Steam and RCS.

2.2 Description of the Proposed Change The proposed change will revise Functional Units 17.A and 17.B of TS Table 4.3-1, "Reactor Trip System Instrumentation Surveillance Requirements". The "S/U" frequency in the TADOT column of this table will be deleted and replaced with a reference to Table Notation (8) which would state: Prior to entering MODE 1 whenever the unit has been in MODE 3. The proposed surveillance frequency does not alter the current provision that takes credit for a surveillance performance within the previous 31 days.

The current surveillance frequency for the turbine trip functions is prior to reactor startup. This requirement is associated with a MODE 1 applicability for these functions. Performance of this test with low heat prior to reactor startup with the MSIVs open challenges plant safety due to the large cooldown incurred. Performing this test prior to reactor startup carries a significant risk of causing a SI due to the rapid cooldown and depressurization of the Main Steam and RCS. With

Document Control Desk RC-18-0020 CR-17-04110 Page 3 of 8 the proposed changes, testing prior to MODE 1 ensures these functions will be OPERABLE when required. These functions can be tested with sufficient nuclear heat available to perform the surveillance without challenging plant safety.

Additionally, this change will align the surveillance requirements and the mode requirement for the Turbine Trip TADOT. There are currently conflicting requirements between the surveillance frequency and the actual mode requirement for the Turbine Trip TADOT. While TS Table 4.3-1 lists the TADOT frequency as required to be performed prior to reactor startup (MODE 3), it lists actual mode for which the surveillance is required as MODE 1. In summary, the surveillance frequency for the turbine trip functions of the reactor trip system instrumentation requirements will be revised to be consistent with the mode applicability for these functions.

The proposed TS changes are noted on the marked-up TS page provided in Attachment 2. The proposed retyped TS are provided in Attachment 3.

3 TECHNICAL EVALUATION The reactor trip on turbine trip function anticipates the loss of heat removal capabilities of the secondary system following a turbine trip. This trip function acts to minimize the pressure/temperature transient on the reactor. The unit is designed to withstand a complete loss of load and not sustain core damage or challenge the RCS pressure limitations. Core protection is provided by the Pressurizer Pressure - High trip Function and RCS integrity is ensured by the pressurizer safety valves. Below the P-9 setpoint, a load rejection can be accommodated by the Steam Dump System. In MODE 2, 3, 4, 5, or 6, there is no potential for a load rejection, and the function does not need to be OPERABLE.

The purpose of the TADOT of the Turbine Trip Function of the Reactor Trip System is to demonstrate operability of the Turbine Stop Valve closure interlock and alarm and the Turbine low emergency trip fluid pressure switches. These trip functions are only enabled by design when the reactor power is above the P-9 interlock that is set at 50 percent rated thermal power.

The proposed change to the turbine trip function surveillance frequency to before entering MODE 1 achieves the appropriate requirement to ensure an operable trip function prior to being required to be operable. The change also ensures an operable trip function prior to the system being capable of initiating the required reactor trip signal.

The turbine trips are not credited as the primary actuation of a reactor trip for any VCSNS postulated event. They are provided to enhance the overall reliability of the reactor protection system. The incorporation of this revision into the VCSNS TSs will not adversely affect the ability of the reactor trip system to generate the required trip signal for turbine trip events.

Document Control Desk RC-18-0020 CR-17-04110 Page 4 of 8 4

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance documents are applicable to the proposed change.

The regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36 "Technical specifications," establish the requirements related to the content of the TS. Section 50.36(c)(3) states:

Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The regulatory requirements in 10 CFR 50.36 are not specific regarding the surveillance requirements. The proposed change assures that the necessary quality of systems and components is maintained, the facility operation will be within safety limits, and that the limiting conditions for operation are met. Therefore, the proposed change is consistent with the requirements of 10 CFR 50.36.

The construction permit for V.C. Summer Unit 1 was issued by the Atomic Energy Commission (AEC) on March 21, 1973. The Operating License was issued on August 6, 1982. NUREG 0717, "Safety Evaluation Report related to the operation of Virgil C. Summer Nuclear Station",

discusses V.C. Summer's conformance with the General Design Criteria. In a letter dated November 14, 1980, V.C. Summer addressed compliance with 10 CFR Parts 20, 50, and 100 including the General Design Criteria. The NRC evaluated the final design and the design criteria and concluded, subject to the applicant's adoption of the additional requirements imposed by the NRC as discussed in the Safety Evaluation Report, that the facility had been designed to meet the requirements of the General Design Criteria. In the November 14, 1980 letter, V.C. Summer provided a discussion to compare the plant design with the General Design Criteria (GDC) as they appeared in 10 CFR 50 Appendix A. It was this discussion, including the identified exceptions, which formed the original plant licensing basis for compliance with the GDC. This discussion is contained in the UFSAR Section 3.0, "General Design Criteria," with more details provided in other UFSAR sections. Changes have been made to the original UFSAR GDC discussions to reflect commitments and changes made to the facility over the life of the plant. Therefore, the GDC discussions in the UFSAR constitute the V.C. Summer Unit 1 licensing bases with respect to compliance with the GDC. These criteria are referenced in Chapter 3.0 of the V.C. Summer UFSAR.

Criterion 20 - Protection System Functions. The protection system shall be designed:

(1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences; and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

The proposed change does not alter the protection system functions. Criterion 20 of the GDC is not affected by the proposed change.

Document Control Desk RC-18-0020 CR-17-04110 Page 5 of 8 Criterion 21 - Protection System Reliability and Testability. The Protection System shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the Protection System shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The Protection System shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

The proposed change alters the surveillance frequency but does not alter the protection system reliability and testability. Criterion 21 of the GDC is not affected by the proposed change.

Criterion 22 - Protection System Independence. The Protection System shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

The proposed change does not alter the design of the protection system and therefore, does not affect the protection system independence. Criterion 22 of the GDC is not affected by the proposed change.

Criterion 23 - Protection System Failure Modes. The Protection System shall be designed to fall into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

The proposed change does not alter the design of the protection system and therefore, does not affect the protection system failure modes. Criterion 23 of the GDC is not affected by the proposed change.

Criterion 24 - Separation of Protection and Control Systems. The Protection System shall be separated from control systems to the extent that failure of a single control system component or channel, or failure or removal from service of any single Protection System component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

Document Control Desk RC-18-0020 CR-17-04110 Page 6 of 8 The proposed change does not alter the design of the protection system and therefore, does not affect the separation of the protection system. Criterion 24 of the GDC is not affected by the proposed change.

Criterion 25 - Protection System Requirements for Reactivity Control Malfunctions. The Protection System shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

The proposed change does not alter the design of the protection system and therefore, Criterion 25 of the GDC is not affected by the proposed change.

Criterion 29 - Protection Against Anticipated Operational Occurrences. The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

The turbine trip functional unit provides anticipatory trips. However, the proposed change does not alter the design of the protection system and therefore Criterion 29 of the GDC is not affected by the proposed change.

The proposed change does not affect plant compliance with these GDC and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Precedent Technical Specification Task Force Traveler TSTF-311, Revision 0, "Revision of Surveillance Frequency for TADOT on Turbine Trip Functional Unit" (ML040620175) was incorporated into Revision 2 of NUREG-1431. Sequoyah Nuclear Plant made a similar change with "Technical Specification Change 07-03 "Revision of Channel Functional Test Surveillance Frequency for Reactor Trip Systems (RTS) Turbine Trip" (ML07108020240)." VCSNS is not based on Improved Technical Specifications. The primary differences between VCSNS TS and NUREG-1431 is that NUREG-1431 allows for:

• The tests to be performed prior to exceeding the [P-9] interlock whenever the unit has been in MODE 3, if not performed within the previous 31 days.

• The NUREG-1431 Action Statement also allows for a reduction to thermal power to below P-9 within 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> if not performed.

4.3 No Significant Hazards Considerations Analysis

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

Response: No. The proposed changes revise the surveillance frequency for reactor trip functions from a turbine trip event. These changes do not alter these functions physically, or how they are maintained. Changing the surveillance from "prior to Startup" to "prior to entering MODE 1" will continue to ensure operability of the function before the plant is in a condition that would benefit from the associated actuation and prior to applicability. Since these changes will

Document Control Desk RC-18-0020 CR-17-04110 Page 7 of 8 not affect the ability of these trips to perform the initiation of reactor trips when appropriate, the offsite dose consequences for an accident will not be impacted. Equally, the potential to cause an accident is not affected because no plant system or component has been altered by the proposed changes.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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 changes only affect surveillance frequency requirements for the turbine trip functions. This does not affect any physical features of the plant, or the manner in which these functions are utilized. The proposed surveillance frequency will require the functions to be verified operable before the turbine trip functions are applicable and able to perform their trip functions. Changing the surveillance from "prior to Startup" to "prior to entering MODE 1" will continue to ensure operability of the function before the plant is in a condition that would benefit from the associated actuation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No. The proposed changes do not alter any plant setpoints or functions that are assumed to actuate in the event of postulated accidents. The proposed changes do not alter any plant feature and only alters the MODE which the surveillance tests must be performed.

The proposed changes ensure the functionality of the turbine trips when assumed in the analysis for accident mitigation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, VCSNS concludes that the proposed amendment does not involve a 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.

4.4 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 Commission's 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.

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment 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 amendment does not involve (1) a significant hazards consideration, (2) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in individual or cumulative occupational radiation

Document Control Desk RC-18-0020 CR-17-04110 Page 8 of 8 exposure. Accordingly, the proposed amendment 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 amendment.

6 REFERENCES

1) U.S. Code of Federal Regulations, Appendix A to Title 10 Part 50 "General Design Criteria"

2) TSTF-311 Rev. 0 "Revision of Surveillance Frequency for TADOT on Turbine Trip Functional Unit" ADAMS Accession No. ML040620175

3) NUREG-1431 Rev. 4 "Standard Technical Specifications, Westinghouse Plants"

4) V.C. Summer FSAR Sections 3.1 "Conformance with NRC General Design Criteria" and 7.2.1.1.2 "Reactor Trips"

5) Sequoyah Nuclear Plant (SQN) - Units 1 and 2 - Technical Specifications (TS) Change 07-03 "Revision of Channel Functional Test Surveillance Frequency for Reactor Trip System (RTS) Turbine Trip." ADAMS Accession No. ML071080240

Document Control Desk RC-18-0020 CR-17-04110 Page 1 of 3 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 2 EXISTING TS PAGES MARKED TO SHOW THE PROPOSED CHANGES Proposed Technical Specification Changes Summary Paqe Affected Section Bar#

Description of Change 3/4 3-12 Table 4.3-1 Items 17.A and 17.B "Trip Actuating Device Operational Test" Replaces the requirement that the TADOT be performed prior to plant startup with the requirement that the TADOT be performed prior to entering MODE 1 whenever the unit has been in MODE 3. The proposed surveillance frequency does not alter the current provision that takes credit for a surveillance performance within the previous 31 days.

3/4 3-14 Table 4.3-1 "Table Notation" Changes Note #8 to, "Prior to entering Mode 1 whenever the unit has been in MODE 3"

(t) c m

73 c

z H

CO CO hJ 3

CD Cl 3

0 3

O CD TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT TRIP ANALOG ACTUATING CHANNEL DEVICE CHANNEL CHANNEL OPERATIONAL OPERATIONAL CHECK CALIBRATION TEST TEST

13.

Steam Generator Water Level-S Low-Low

14.

Steam Generator Water Level - S Low Coincident with Steam/

Feedwater Flow Mismatch

15.

Undervoltage - Reactor Coolant N.A.

Pumps

16.

Underfrequency - Reactor N.A.

Coolant Pumps

17.

Turbine Trip A. Low Fluid Oil Pressure N.A.

B. Turbine Stop Valve N.A.

Closure

19.

Reactor Trip System Interlocks A. Intermediate Range N.A.

Neutron Flux, P-6 B. Low Power Reactor N.A.

Trips Block, P-7 C

Power Range Neutron N.A.

Flux, P-8 R

R R

R R

R R(4)

R(4)

R(4)

SA SA N.A.

N.A.

N.A.

N.A.

R R

R N.A.

N.A.

SA SA (1,8,10)

ACTUATION LOGIC TEST N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

(1,8,10)

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

MODES FOR WHICH SURVEILLANCE IS REQUIRED 1,2 1,2 1

1 1

1 2##

1 1

TABLE 4.3-1 (Continued)

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

Below P-6 (Intermediate Range Neutron Flux Interlock) setpoint.

Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) setpoint.

(1)

If not performed in previous 31 days.

(2)

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2 percent. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(3)

Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3 percent. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

Detector plateau curves shall be obtained evaluated and compared to manufacturer's data. For the Power Range Neutron Flux Channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7)

Each train shall be tested at least every 124 days on a STAGGERED TEST BASIS.

(8)

- DELETED Prior to entering MODE 1 whenever the unit has been in MODE 3.

(9)

Surveillance in MODES 3*, 4* and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(10) -

Setpoint verification is not required.

(11)-

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(12) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(13) -

Local manual shunt trip prior to placing breaker in service.

(14) -

Automatic undervoltage trip.

(15) -

Each train shall be tested at least every 184 days on a Staggered Test Basis.

(16) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10 and 184 days thereafter.

(17) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-6 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3 from MODE 2 and 184 days thereafter.

(18)

If not performed in previous 184 days.

SUMMER - UNIT 1 3/4 3-14 Amendment No. 73, 78, 101, 20©

Document Control Desk RC-18-0020 CR-17-04110 Page 1 of 3 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 3 REVISED (CLEAN) TS PAGES Replace the following pages of the Technical Specifications with the attached revised pages.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages 3/4 3-12 3/4 3-14 Insert Pages 3/4 3-12 3/4 3-14

(f) c m

73 TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CO CO

..A ro 3

0 3

Q.

3 0

3 i+* z o

FUNCTIONAL UNIT TRIP ANALOG ACTUATING CHANNEL DEVICE CHANNEL CHANNEL OPERATIONAL OPERATIONAL CHECK CALIBRATION TEST TEST

13.

Steam Generator Water Level-S Low-Low

14.

Steam Generator Water Level -

S Low Coincident with Steam/

Feedwater Flow Mismatch

15.

Undervoltage - Reactor Coolant N.A.

Pumps

16.

Underfrequency - Reactor N.A.

Coolant Pumps

17.

Turbine Trip A. Low Fluid Oil Pressure N.A.

B. Turbine Stop Valve N.A.

Closure

19.

Reactor Trip System Interlocks A. Intermediate Range N.A.

Neutron Flux, P-6 B. Low Power Reactor N.A.

Trips Block, P-7 C

Power Range Neutron N.A.

Flux, P-8 R

R R

R R

R R(4)

R(4)

R(4)

SA SA N.A.

N.A.

N.A.

N.A.

R R

R N.A.

N.A.

SA SA (1,8,10)

(1,8,10)

N.A.

N.A.

N.A.

ACTUATION LOGIC TEST N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

MODES FOR WHICH SURVEILLANCE IS REQUIRED 1,2 1,2 1

1 1

1 2##

1 1

TABLE 4.3-1 (Continued)

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

Below P-6 (Intermediate Range Neutron Flux Interlock) setpoint.

Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) setpoint.

(1)

If not performed in previous 31 days.

(2)

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2 percent. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(3)

Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3 percent. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

Detector plateau curves shall be obtained evaluated and compared to manufacturer's data. For the Power Range Neutron Flux Channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7)

Each train shall be tested at least every 124 days on a STAGGERED TEST BASIS.

(8)

Prior to entering MODE 1 whenever the unit has been in MODE 3.

(9)

Surveillance in MODES 3*, 4* and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(10) -

Setpoint verification is not required.

(11) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(12) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(13) -

Local manual shunt trip prior to placing breaker in service.

(14) -

Automatic undervoltage trip.

(15) -

Each train shall be tested at least every 184 days on a Staggered Test Basis.

(16) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10 and 184 days thereafter.

(17) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-6 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3 from MODE 2 and 184 days thereafter.

(18)

If not performed in previous 184 days.

SUMMER-UNIT 1 3/4 3-14 Amendment No. 73,78,101,209