RC-18-0113, (Vcsns), Unit 1 - Supplement to License Amendment Request LAR-18-03422 Request for a One-Time Extension to the Surveillance Frequency 4.3.3.6 of the Core Exit Temperature Instrumentation

(Vcsns), Unit 1 - Supplement to License Amendment Request LAR-18-03422 Request for a One-Time Extension to the Surveillance Frequency 4.3.3.6 of the Core Exit Temperature Instrumentation

ML18243A392
Person / Time
Site:	Summer
Issue date:	08/31/2018
From:	Lippard G South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR-18-03422, RC-18-0113
Download: ML18243A392 (17)

Text

George A. Lippard Vice President, Nuclear Operations 803.345.4810 A SCANA COMPANY August 31, 2018 RC-18-0113 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 SUPPLEMENT TO LICENSE AMENDMENT REQUEST LAR-18-03422 REQUEST FOR A ONE-TIME EXTENSION TO THE SURVEILLANCE FREQUENCY 4.3.3.6 OF THE CORE EXIT TEMPERATURE INSTRUMENTATION

References:

1. Letter from George A. Lippard to NRC Document Control Desk dated August 24, 2018. "LICENSE AMENDMENT REQUEST LAR-18-03422, REQUEST FOR A ONE-TIME EXTENSION TO THE SURVEILLANCE FREQUENCY

.4.3.3.6 OF THE CORE EXIT TEMPERATURE INSTRUMENTATION"

[ML18236A383]

On August 24, 2018 VCSNS submitted a License Amendment Request, pursuant to 10 CFR 50.90 for a one-time license amendment change to the Technical Specification (TS) surveillance requirement of the Core Exit Temperature Instrumentation channel calibrations (Reference 1).

The requested amendment proposed a change to the surveillance requirement of TS 4.3.3.6 to allow a one-time extension of the frequency of the Core Exit Temperature Instrumentation Channel Calibrations from "every refueling outage", which is interpreted as every 18 months, to 19 months.

VCSNS is requesting this license amendment request to be processed as an exigent TS Change to expedite the review and approval process by September 24, 2018. The issuance of the Safety Evaluation by September 24, 2018 is requested to support the Core Exit Thermocouple TS Surveillance channel calibration, currently scheduled to begin on September 25, 2018. This work typically requires three shifts and will encroach upon the 25% grace period for this surveillance, which ends September 28, 2018.

Therefore, pursuant to 10 CFR 50.90 and10 CFR 50.91(a)(6), South Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for South Carolina Public Service authority is resubmitting the license amendment request as an exigent request. Attachment 1 has been V. C. Summer Nuclear Station

• P. 0. Box 88

• Jenkinsville, South Carolina

• 29065

• F (803) 941-9776

• www.sceg.com

Document Control Desk RC-18-0113 CR-18-03422 Page 2 of 3 updated to explain the exigency and why it could not be avoided per the requirement of 10 CFR 50.91(a)(6).

Additionally, an oversight was discovered in the original submission and has been corrected in . The date the decision was made to not perform dewatering of the thimbles online was corrected from August 8th, 2018 to August 7th, 2018. This oversight is documented in CR-18-03555. The remaining attachments are unchanged but included in this supplement for completeness. contains the marked-up version of the affected TS page. Attachment 3 contains the reprinted version of the affected TS pages. Attachment 4 is the new commitment page.

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.

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).

Approval of the proposed amendment is requested by September 24, 2018. Once approved, the amendment shall be implemented immediately and will remain in effect until the conclusion of Refueling Outage 24.

The proposed change does introduce one new commitment. If there are any questions or if additional information is needed, please contact Michael S. Moore at (803) 345-4752.

Document Control Desk RC-18-0113 CR-18-03422 Page 3 of 3 declare under penalty of perjury that the foregoing is true and correct.

2 3>l 11 Y Executed on BAB/GAL/wk Attachment 1 Description and Assessment of the Proposed Changes Attachment 2 Existing TS Pages Marked to Show the Proposed Changes Attachment 3 Revised (Clean) TS Pages Attachment 4 List of Regulatory Committments c: J. E. Addison NRC Resident Inspector W. K. Kissam K. M. Sutton J. B. Archie S. E. Jenkins J. H. Hamilton P. Ledbetter G. J. Lindamood NSRC W. M. Cherry RTS (CR-18-03422)

C. Haney File (813.20)

S. A. Williams PRSF (RC-18-0113)

Document Control Desk RC-18-0113 CR-18-03422 Page 1 of 9 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) for a one-time extension of the of the channel calibration frequency of the Core Exit Temperature Instrumentation.

1

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90 and 10 CFR 50.91(a)(6), 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 a one-time extension of TS Surveillance Requirement 4.3.3.6 specifically for the Channel Calibration of Item 12- Core Exit Temperature Instrumentation in TS Table 3.3-

10. The Surveillance Frequency described in TS 4.3.3.6 for the Channel Calibration of Item 12-Core Exit Temperature Instrumentation would receive a one-time extension from every "refueling outage", which is interpreted as every 18 months, to 19 months.

Compliance with the Action Statement for this TS Surveillance could increase the probability of a TS required plant shutdown before the start of a scheduled Refueling Outage (RF24). RF24 is scheduled to start October 6, 2018 and the 25% grace period for the Channel Calibrations ends September 28, 2018.

The concern with performance of the Surveillance, is that if any thermocouple strings were found out of tolerance and needed adjustment, they would be required to be removed from the Power Distribution Monitoring System (PDMS). If more than TS allowable thermocouples are removed, the requirements of TS 3.3.3.11.b.2 would no longer be satisfied and PDMS would become inoperable. With an inoperable PDMS, plant maneuvers during the performance of planned pre-outage surveillances and testing would be performed without power distribution monitoring capability and increase the probability of a TS required plant shutdown. An unscheduled shutdown would also incur numerous scheduling impacts related to contracted vendors and other support services already scheduled to support RF24.

The requested surveillance extension will allow one month more than the currently specified refueling outage interval of 18 months plus four and one half months grace period, which will allow the plant to stay online until the scheduled start of RF24. In order to provide a more

Document Control Desk RC-18-0113 CR-18-03422 Page 2 of 9 specific description of the proposed change, a markup of the TS page is attached and a description of the change with appropriate justification is summarized in this document.

2 DETAILED DESCRIPTION 2.1 System Design and Operation The Incore Temperature Monitoring System is designed to provide rapid monitoring of fuel assembly outlet temperatures and to verify that the core is being adequately cooled (subcooling and natural circulation) during and after an accident.

As a secondary function, the Incore Temperature Monitoring System, is designed to provide an accurate measure of the relative, integrated fuel assembly power distribution. This function requires the use of the reactor core flux maps from the Incore Neutron Flux Monitoring System to normalize the temperature measurements.

The thermocouple temperature data along with reactor coolant system pressure, loop flows, loop inlet temperatures, average of the loop temperatures, loop differential temperatures, and power level is used to determine:

(1) Saturation margin (2) Core relative fuel assembly power distribution (3) Core enthalpy rise nuclear hot-channel factors (4) Core radial tilting factors The Incore Temperature Monitoring System provides isolated analog input signals for sixteen (16) thermocouples (two per core quadrant per thermocouple train) to the Core Cooling Monitoring System. The signals from the Train A and Train B isolator cabinets are routed, maintaining electrical separation and utilizing qualified cable and seismically qualified cable trays and conduit, to the respective Core Cooling Monitors "A" and "B".

The Core Cooling Monitoring System is designed to provide information to plant personnel concerning the status of reactor core heat removal capability. This information includes a continuous display of the saturation margin to provide an early warning that core conditions are approaching saturation. In addition, the core outlet temperatures at selected fuel assemblies are displayed to assist in the diagnosis of inadequate core cooling. The Core Cooling Monitoring System was designed in response to the post Three Mile Island requirements documented by the NRC in NUREG-0578, Item 2.1.3b.

2.2 Current Technical Specifications Requirements The current VCSNS TS require the accident monitoring instrumentation channels shown in Table 3.3-10 to be operable in Modes 1,2, and 3. Item 12 of Table 3.3-10 lists the requirements of the Core Exit Temperature Instrumentation. The "Required Number of Channels" is 4 per core quadrant per channel. The "Minimum Channels Operable" is 2 per core quadrant per channel. The Action statements are listed below:

• With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown on Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 30 days or:

Document Control Desk Attachment 1 RC-18-0113 CR-18-03422 Page 3 of 9 o Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and o Submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

• With the number of OPERABLE accident monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

TS Section 4.3.3.6 describes the Surveillance Requirements of the accident monitoring instrumentation channels shown in Table 3.3-10. It is required that each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every refueling outage.

VCSNS does not utilize Standard Technical Specifications as described in NUREG-1431, Volume 1, Revision 4. NUREG-1431 Revision 4 discusses Post Accident Monitoring (PAM)

Instrumentation in section 3.3.3. Instrumentation is required to be OPERABLE in MODES 1, 2, and 3. CHANNEL CHECKS are required every 31 days or in accordance with the Surveillance Frequency Control Program. CHANNEL CALIBRATIONS are required every 18 months or in accordance with the Surveillance Frequency Control Program. 2 Channels are required to be OPERABLE per quadrant, with each Channel consisting of 2 core exit thermocouples (CETs).

With less than the required number of channels, the required action is to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.3 Reason for the Proposed Change Surveillance test STP-345.047A ,"A-Train In-Core Thermocouple Isolation Amplifier Calibration",

is required to be performed on a every refueling outage frequency per TS 4.3.3.6. The grace period for the surveillance test ends on September 28, 2018. The procedure was started on June 19, 2018, but was not able to be completed due to inadequate fire watch resources at the time. Before the test could be rescheduled, unidentified leakage was noted to be elevated while performing STP-114.002, "Operational Leakage Calculation", on June 21, 2018.

Reactor building entries were made on June 21, 2018 and determined that incore thimble B-7 was leaking around the 'A' 10-path. The leak was isolated by closing XVT58001-IC Fuel Assembly B-7 thimble isolation valve. From the time the leak was isolated on June 21, 2018 to August 2, 2018 the station made ten (10) reactor building entries to determine the status of the incore detector system and attempt to restore the system to operable status.

During system functional testing, it was determined that water in some of the thimble tubes prevented the system from being restored to an operable status.

A decision was made on August 7, 2018 to not perform dewatering of the thimbles online due to the radiological and industrial risk associated with the task. Preparations to restore the system during the RF24 are ongoing.

Document Control Desk RC-18-0113 CR-18-03422 Page 4 of 9 TS 3.3.3.2 action states that with the moveable incore detection system inoperable, do not use the system for the functions listed below:

a) Recalibration of the excore neutron flux detection system, b) Monitoring the QUADRANT POWER TILT RATIO using a full-core flux map per Specification 4.2.4.2, or c) Measurement of FNah and FQ(Z).

TS allows the use of PDMS or the incore detection system to perform these functions. The station has been relying upon the PDMS to monitor core power distribution.

During the performance of Surveillance test STP-345.047A, "A-Train In-Core Thermocouple Isolation Amplifier Calibration", when a core exit thermocouple string is found to be out of tolerance the string is adjusted to bring it back to within the specified calibration tolerance of +/-

0.050mV (+/-2.13 deg °F) The adjustment of the thermocouple string then makes it unavailable for use in PDMS due to the fact that PDMS was calibrated to the previous thermocouple reading.

The concern with performance of the Surveillance, is that if any thermocouple strings were found out of tolerance and needed adjustment, they would be required to be removed from the Power Distribution Monitoring System (PDMS). If more than TS allowable thermocouples are removed, the requirements of TS 3.3.3.11.b.2 would no longer be satisfied and PDMS would become inoperable. With an inoperable PDMS, plant maneuvers during the performance of planned pre-outage surveillances and testing would be performed without power distribution monitoring capability and increase the probability of a TS required plant shutdown. An unscheduled shutdown would also incur numerous scheduling impacts related to contracted vendors and other support services already scheduled to support RF24.

The action statement for PDMS being inoperable is to use the moveable incore detector system to satisfy the requirements of TS 3.3.3.2 to obtain any required power distribution measurements. As stated previously, the incore system is inoperable therefore this action statement cannot be met. The requested surveillance extension will allow one month more than the currently specified refueling outage interval of 18 months plus four and one half months grace period, which will allow the plant to stay online until the scheduled start of RF24.

2.4 Description of the Proposed Change VCSNS TS 4.3.3.6 describes the Surveillance Requirements of Accident Monitoring Instrumentation and currently reads as follows:

"Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every refueling outage. The Reactor Building Radiation Level Instrumentation CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for the range decades above 10 R/hr and a single point calibration of the detector below 10 R/hr with an installed or portable gamma source."

Document Control Desk RC-18-0113 CR-18-03422 Page 5 of 9 A one-time extension, which will expire at the end of Refuel 24, will be added for Item 12 of Table 3.3-10. This will revise the Surveillance Requirement of the Core Exit Temperature Instrumentation by extending the CHANNEL CALIBRATION to 19 Months. The new Surveillance Requirement will read as follows:

"Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every refueling outage, with the exception of item 12 in Table 3.3-

10. Item 12 shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every 19 months. The Reactor Building Radiation Level Instrumentation CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for the range decades above 10 R/hr and a single point calibration of the detector below 10 R/hr with an installed or portable gamma source."

3 TECHNICAL EVALUATION The proposed TS change will not adversely affect the ability of the core exit thermocouples to perform their specified safety functions as described in the system design and operation section of this change request.

Historical review of the last two performances of this surveillance have shown that 6 of 24 and 7 of 24 thermocouples were required to be adjusted back into tolerance due to as found values being out of tolerance. Therefore 18 and 17 thermocouples, respectively, were in tolerance and would have remained operable if no adjustment were performed.

V.C. Summer Core Exit Thermocouple Uncertainty Calculations for Emergency Operating Procedures (EOP) Applications Including Subcooling (Reference 4) conservatively assumes a drift value of +/-1.0°F per six month period for the thermocouples. Industry standards typically use

+/-1.0°F per year as documented in Process/Industrial Instruments and Controls Handbook, Fifth Edition for thermocouples. The calculation also applies a drift term for the Analog to Digital Isolation Amplifier of +/-0.1% per six months which equates to a drift value of +/-2.13°F per six month period for this component. An allowance of +/-6.0°F was also applied to this calculation for the drift and temperature input uncertainties of the Core Cooling Monitor. This allowance is inclusive of the temperature and drift effect on both the thermocouples and wide range pressure input. The drift term for the remaining components associated with the core exit thermocouple uncertainty calculation were evaluated for a 24 month frequency.

The additional drift that the core exit thermocouples would incur during the eight day time period between when the surveillance grace period ends on September 28, 2018 and when the plant is scheduled to exit the TS applicability conditions (Enter Mode 4) on October 6, 2018 is minor in comparison to the uncertainty allowances associated with subcooling and temperature monitoring.

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

Document Control Desk Attachment 1 RC-18-0113 CR-18-03422 Page 6 of 9 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.

10 CFR 50.36 sets the regulatory requirements for the content of the Technical Specifications. As quoted above, 10 CFR 50.36(c)(3) of this regulation requires, in part, that the plant Technical Specifications contain surveillance requirements. The proposed change to SR 4.3.3.6 is for a short duration, one-time extension of its surveillance frequency, which is not specified in the regulation.

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.

VCSNS meets the intent of Regulatory Guide 1.97 Revision 3, "Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident".

Regulatory Guide 1.97 details the applicable GDCs as discussed below:

Criterion 13-Instrumentation and Control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

The proposed change does not alter Criterion 13 for Instrumentation and Control. Movable incore flux detectors and fixed incore thermocouples are provided as operational aids to the operator. FSAR Chapter 7 contains further details on instrumentation and controls. FSAR

Document Control Desk RC-18-0113 CR-18-03422 Page 7 of 9 Section 7.5 details the information available to the operator for the performance of required safety functions. These systems are also described in FSAR Chapters 6, 7, 8, 9, 11 and 12.

Criterion 19 -Control Room. The Control Room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

The proposed change does not alter the Control Room. Criterion 19 of the GDC is not affected by the proposed change.

Criterion 64- Monitoring Radioactivity Releases. Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

The proposed change does not alter the Monitoring Radioactivity Releases. Criterion 64 of the GDC is not affected by the proposed change.

VCSNS will continue to meet the intent of Regulatory Guide 1.97, Revision 3 Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following An Accident (REVISION 3; 12/83).

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

10 CFR 50.91 (a)(6)(vi) states that "Where the Commission finds that exigent circumstances exist, in that a licensee and the Commission must act quickly and that time does not permit the Commission to publish a Federal Register notice allowing 30 days for prior public comment, and it also determines that the amendment involves no significant hazards considerations, it:

(vi) Will require the licensee to explain the exigency and why the licensee cannot avoid it, and use its normal public notice and comment procedures in paragraph (a)(2) of this section if it determines that the licensee has failed to use its best efforts to make a timely

Document Control Desk Attachment 1 RC-18-0113 CR-18-03422 Page 8 of 9 application for the amendment in order to create the exigency and to take advantage of this procedure.

From the time the leak was isolated on June 21, 2018 to August 2, 2018 there have been 10 reactor building entries made to determine the status of the incore detector system and attempt to restore the system to operable status. The decision was made on August 7th, during the bimonthly monthly meeting of the Plant Health Committee (PHC), to not perform dewatering of the thimbles online due to the radiological and industrial risk associated with the task. Following the PHC meeting, VCSNS began formulating and discussing possible solutions. On August 15th, it was decided that the best solution would be to request a license amendment to delay the channel calibration. NRC Headquarters was contacted on August 16, 2018 to discuss the license amendment request. VCSNS submitted this change (Reference 1) on August 24, 2018 to allow for a 30 day public comment period. Due to the impracticality of issuing the Safety Evaluation by the station need date of September 24, 2018, the license amendment request is now being resubmitted as an exigent change.

4.2 Precedent No precedent has been found for a one-time extension of the surveillance frequency of Core Exit Temperature Instrumentation.

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 change is a short duration, one-time extension to the surveillance frequency requirement of channel calibrations of the Core Exit Temperature Instrumentation. The performance of the surveillance, or the failure to perform the surveillance, is not a precursor to an accident. An extension in performing the surveillance does not result in the system being unable to perform its function. The systems required to mitigate accidents will remain capable of performing their required functions. No new failure modes have been introduced because of this action and the consequences remain consistent with previously evaluated accidents.

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 change only affects the surveillance frequency requirement for the channel calibrations of the Core Exit Temperature Instrumentation.

This proposed change does not involve a change to any physical features of the plant, or the manner in which these functions are utilized. No few failure mechanisms will be introduced by the one-time surveillance extension being requested.

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

Document Control Desk Attachment 1 RC-18-0113 CR-18-03422 Page 9 of 9

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

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

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

3) V.C. Summer FSAR Sections 3.1 "Conformance with NRC General Design Criteria" and 7.5 "Safety Related Display Instrumentation"

4) Calculation CN-SCS-05-4 Revision 0, "V.C Summer (CGE) Core Exit Thermocouple Uncertainty Calculations for EOP Applications Including Subcooling.

Document Control Desk RC-18-0113 CR-18-03422 Page 1 of 2 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 2 EXISTING TS PAGES MARKED TO SHOW THE PROPOSED CHANGES Proposed Technical Specification Changes Summary

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown on Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 30 days or submit a Special Report within the following 14 days from the time the action is required. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to operable status.

b.1 With the number of OPERABLE Reactor Building radiation monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

i) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and ii) Submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

b.2 Deleted b.3 With the number of OPERABLE accident monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every refueling outage. The Reactor Building Radiation Level Instrumentation CHANNEL CALIBRATION may aonsist of an electronic calibration of the channel, not including the detector, for the range decades above 10R/hr and a single point calibration of the detector below 10R/hr with an installed or portable gamma source.

Insert:

, with the exception of item 12 in Table 3.3-10. Item 12 shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every 19 months.

SUMMER - UNIT 1 3/4 3-56 Amendment No. 118, 170,

Document Control Desk RC-18-0113 CR-18-03422 Page 1 of 2 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 Insert Pages 3/4 3-56 3/4 3-56

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown on Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 30 days or submit a Special Report within the following 14 days from the time the action is required. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to operable status.

b.1 With the number of OPERABLE Reactor Building radiation monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

i) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and ii) Submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

b.2 Deleted b.3 With the number of OPERABLE accident monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every refueling outage, with the exception of item 12 in Table 3.3-10. Item 12 shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every 19 months. The Reactor Building Radiation Level Instrumentation CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for the range decades above 10R/hr and a single point calibration of the detector below 10R/hr with an installed or portable gamma source.

SUMMER - UNIT 1 3/4 3-56 Amendment No. 118, 170,

Document Control Desk Attachment 4 RC-18-0113 CR-18-03422 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 4 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by the Virgil C. Summer Nuclear Station (VCSNS) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Michael S. Moore at (803) 345-4752.

COMMITMENT DUE DATE VCSNS will submit a License Amendment Request to remove the revision to This is due to NRC within 30 days of the Technical Specification 4.3.3.6 detailed in completion of RF-24.

this LAR.