ML21280A328

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Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions
ML21280A328
Person / Time
Site: Millstone, Summer, North Anna  Dominion icon.png
Issue date: 10/07/2021
From: Mark D. Sartain
Dominion Energy Nuclear Connecticut, Dominion Energy Services, Dominion Energy South Carolina, Virginia Electric & Power Co (VEPCO)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
21-146
Download: ML21280A328 (41)
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Text

{{#Wiki_filter:Dominion Energy Services, Inc. 5000 Dominion Boulevard, Glen Allen, VA 23060 DominionEnergy.com U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 October 7, 2021 DOMINION ENERGY NUCLEAR CONNECTICUT, INC. ~ Dominion ~ Energy Serial No.: NRA/CLT Docket Nos.: License Nos. 10 CFR 50.90 21-146 RO 50-423 50-338/339 50-395 NPF-49 NPF-4/7 NPF-12 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA) DOMINION ENERGY SOUTH CAROLINA, INC. MILLSTONE POWER STATION UNIT 3 NORTH ANNA POWER STATION UNITS 1 AND 2 V. C. SUMMER NUCLEAR STATION UNIT 1 APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-569, "REVISION OF RESPONSE TIME TESTING DEFINITIONS" Pursuant to 10 CFR 50.90, Dominion Energy Connecticut, Inc., Dominion Energy Virginia, and Dominion Energy South Carolina, Inc., (hereafter "the Licensees") are submitting a request for amendments to the Technical Specifications (TS) for Millstone Power Station (MPS) Unit 3, North Anna Power Station (NAPS) Units 1 and 2, and V. C. Summer Nuclear Station (VCS) Unit 1, respectively. The licensees request adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the MPS Unit 3, NAPS Units 1 and 2, and VCS Unit 1 TS. The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time Testing, as applicable. The enclosure provides a description and assessment of the proposed changes. provides marked-up MPS Unit 3, NAPS Units 1 and 2, and VCS Unit 1 TS pages showing the proposed changes. Attachment 2 provides revised (clean) TS pages for the three stations. Attachment 3 provides existing TS Bases pages for the three stations marked-up to show the proposed changes for information only. The Facility Safety Review Committee at each station has reviewed and concurred with the determinations herein. Approval of the proposed amendment is requested by September 30, 2022 Once approved, the amendment shall be implemented within 30 days. This letter contains no regulatory commitments.

Serial No. 21-146 Docket Nos. 50-423/338/339/395 Page 2 of 4 In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State Officials. If you have any questions regarding this submittal, please contact Ms. Cathie Tiernan at (804) 273-3090. Sincerely, ~{~ Mark D. Sartain Vice-President - Nuclear Engineering and Fleet Support

Enclosure:

Description and Assessment Attachments:

1. Proposed Technical Specifications Changes (Mark-Up)
2. Revised Technical Specifications Pages
3. Proposed Technical Specification Bases Changes (Mark-Up)

COMMONWEAL TH OF VIRGINIA ) ) COUNTY OF HENRICO ) The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc., Virginia Electric and Power Company, and Dominion Energy South Carolina, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief. Acknowledged before me this 1+\\\\ day of OcJoh-e'V' My Commission Expires:

!Cl:(\\, 311 '202'-{

Kathryn Hill Barret Notary Public Commonwealth of Virginia M . Reg. No. 7905256 y Commission Expires January 31, 2024 , 2021 ~411~11~ Notary Public

cc: U. S. Nuclear Regulatory Commission, Region I Regional Administrator 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 U.S. Nuclear Regulatory Commission, Region II Regional Administrator Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRG Senior Resident Inspector Millstone Power Station NRG Senior Resident Inspector North Anna Power Station NRG Senior Resident Inspector V.C. Summer Nuclear Station Mr. G. Edward Miller Serial No. 21-146 Docket Nos. 50-423/338/339/395 Page 3 of 4 NRC Senior Project Manager - North Anna Units 1 and 2 and V. C. Summer U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 Mr. R. V. Guzman NRG Senior Project Manager - Millstone Units 2 and 3 U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 C-2 11555 Rockville Pike Rockville, MD 20852-2738 Old Dominion Electric Cooperative Electronically Distributed Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

State Health Commissioner Virginia Department of Health James Madison Building - 7th Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Ms. Anuradha Nair-Gimmi Bureau of Environmental Health Services Serial No. 21-146 Docket Nos. 50-423/338/339/395 Page 4 of 4 South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Mr. G. J. Lindamood Santee Cooper-Nuclear Coordinator 1 Riverwood Drive Moncks Corner, SC 29461

Enclosure DESCRIPTION AND ASSESSMENT Dominion Energy Connecticut, Inc. (DENC) Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC) Millstone Power Station Unit 3 North Anna Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

1.0 DESCRIPTION

Serial No. 21-146 Docket Nos. 50-423/338/339/395 Adoption of TSTF-569 Enclosure Page 2 of 7 DESCRIPTION AND ASSESSMENT Pursuant to 10 CFR 50.90, Dominion Energy Connecticut, Inc., (DENG), Virginia Electric and Power Company (Dominion Energy Virginia), and Dominion Energy South Carolina, Inc. (DESC), (hereafter "the Licensees") request adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Millstone Power Station (MPS) Unit 3, North Anna Power Station (NAPS) Units 1 and 2, and V. C. Summer Nuclear Station (VCS) Unit 1 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 The licensees have 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. The licensees have concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to MPS Unit 3, NAPS Units 1 and 2, and VCS Unit 1 and justify this amendment for the incorporation of the changes into the respective plants' TS. 2.2 Variations MPS is proposing the following two variations from the TS changes described in TSTF-569 or the applicable parts of the NRC staff safety evaluation dated August 14, 2019.

1. TSTF-569 and the NRC staff safety evaluation refer to the 1977 and 1987 versions of IEEE Standard 338, "Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems," and the 2012 version titled, "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems." In the safety evaluation, the NRC staff found that replacement components will continue to perform the same design functions as the original components.

Furthermore, the NRC staff found that the methodologies contained in TSTF-569, Attachment 1, provide adequate criteria for ensuring that replacement component's degraded response time issues or failures would be captured, and, as a result, conformance with the IEEE 338 design criteria is not affected. MPS Unit 3 is committed to the 1971 version of IEEE Standard 338, which does not discuss response time testing. However, response time testing is required by the MPS TS. The staffs conclusion is equally applicable

Serial No. 21-146 Docket Nos. 50-423/338/339/395 Adoption of TSTF-569 Enclosure Page 3 of 7 to the MPS implementation of TSTF-569 as the 1971 design criteria are not affected.

2. MPS Unit 3 TS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the MPS Unit 3 TS for RTS Instrumentation is 3.3.1, whereas the TSTF-569 numbering for RTS Instrumentation is 3.3.1.16. Also, the MPS Unit 3 TS for ESFAS Instrumentation is 3.3.2, whereas the TSTF-569 numbering for ESFAS Instrumentation is 3.3.10. This difference is administrative and does not affect the applicability of TSTF-569 to the MPS Unit 3 TS. Finally, the MPS Unit 3 ESF RTT definition uses the phrase "Engineered Salety Features" vs. "Engineered Safety Feature" in the TSTF-567 ESF RTT definition. This minor difference has no effect on the requested TS change.

NAPS is proposing the following two variations from the TS changes described in the TSTF-569 or the applicable parts of the NRC staff safety evaluation dated August 14, 2019.

1. TSTF-569 and the NRC staff safety evaluation refer to the 1977 and 1987 versions of IEEE Standard 338, "Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems," and the 2012 version titled, "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems." In the safety evaluation, the NRC staff found that replacement components will continue to perform the same design functions as the original components.

Furthermore, the NRC staff found that the methodologies contained in TSTF-569, Attachment 1, provide adequate criteria for ensuring that replacement components degraded response time issues or failures would be captured, and, as a result, conformance with the IEEE 338 design criteria is not affected. NAPS Units 1 and 2 are committed to the 1971 version of IEEE Standard 338, which does not discuss response time testing. However, response time testing is required by the NAPS TS. The staffs conclusion is equally applicable to the NAPS implementation of TSTF-569 as the 1971 design criteria are not affected.

2. NAPS TS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the NAPS TS for ESF Actuation System (ESFAS) Instrumentation is 3.3.9, whereas the TSTF-569 numbering for ESFAS Instrumentation is 3.3.10. This difference is administrative and does not affect the applicability of TSTF-569 to the NAPS Units 1 and 2 TS.

VCS is proposing the following four variations from the TS changes described in TSTF-569 or the applicable parts of the NRC staff safety evaluation dated August 14, 2019.

Serial No. 21-146 Docket Nos. 50-423/338/339/395 Adoption of TSTF-569 Enclosure Page4 of 7

1. TSTF-569 and the NRG staff safety evaluation refer to the 1977 and 1987 versions of IEEE Standard 338, "Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems," and the 2012 version titled, "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems." In the safety evaluation, the NRG staff found that replacement components will continue to perform the same design functions as the original components.

Furthermore, the NRG staff found that the methodologies contained in TSTF-569, Attachment 1, provide adequate criteria for ensuring that replacement components' degraded response time issues or failures would be captured, and, as a result, conformance with the IEEE 338 design criteria is not affected. VCS Unit 1 is committed to the 1971 version of IEEE Standard 338, which does not discuss response time testing. However, response time testing is required by the VCS Unit 1 TS. The staffs conclusion is equally applicable to the VCS implementation of TSTF-569 as the 1971 design criteria are not affected.

2. VCS Unit 1 TS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the VCS Unit 1 TS for RTS Instrumentation is 3.3.1, whereas the TSTF-569 numbering for RTS Instrumentation is 3.3.10. Additionally, the VCS Unit 1 TS for ESFAS Instrumentation is 3.3.2, whereas the TSTF-569 numbering for ESFAS Instrumentation is 3.3.10. This difference is administrative and does not affect the applicability of TSTF-569 to the VCS TS.
3. VCS Unit 1 TS Bases utilize a different reference format than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the VCS Unit 1 TS Bases do not itemize references in each bases section. Therefore, the TS Bases change identifies the documents rather than noting a reference. This difference is administrative and does not affect the applicability of TSTF-569 to the VCS Unit 1 TS.
4. VCS Unit 1 TS Amendment 158 approved two additional upgraded 7300 process cards, for which WCAP-14036 was applicable. The VCS Unit 1 TS Bases were revised to include these additional cards.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination The licensees request adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Millstone Power Station (MPS) Unit 3, North Anna Power Station (NAPS) Units 1

Serial No. 21-146 Docket Nos. 50-423/338/339/395 Adoption of TSTF-569 Enclosure Page 5 of 7 and 2, and V. G. Summer Nuclear Station (VGS) Unit 1 Technical Specifications (TS). The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF} Response Time and Reactor Trip System (RTS) Response Time. The licensees have evaluated whether a significant hazards consideration is involved with the proposed amendment(s} by focusing on the three standards set forth in 10 GFR 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 licensees to evaluate using an NRG-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 NRG-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 licensees to evaluate using an NRG-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

Serial No. 21-146 Docket Nos. 50-423/338/339/395 Adoption of TSTF-569 Enclosure Page 6 of 7 assumptions made 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 licensees to evaluate using an NRG-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, the licensees conclude 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 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.

4.0 EVIRONMENT AL EVALUATION Serial No. 21-146 Docket Nos. 50-423/338/339/395 Adoption of TSTF-569 Enclosure Page 7 of 7 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 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. TECHNICAL SPECIFICATIONS PROPOSED CHANGES Dominion Energy Connecticut, Inc. (DENC) Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC) Millstone Power Station Unit 3 North Anna Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microCurie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131 m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III. I of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." 1.12 DELETED ENGINEERED SAFETY FEATURES RESPONSE TIME l.13 The ENGINEERED SAFETY FEATURES (ESF) 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. 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 the methodology for verification have been previously reviewed and approved by the NRC. 1.14 DELETED FREQUENCY NOTATION , or the components have been evaluated in accordance with an NRG approved methodology 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. LEAKAGE 1.16 LEAKAGE shall be: 1.16.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or MILLSTONE - UNIT 3 1-3 Amendment No. -84,.s:7-, --1-U, -l-81, ~.

m,~.~

August l '.2, ;!998 DEFINITIONS PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. 1.22 DELETED PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3650 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.28 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. 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 the methodology for verification have been previously reviewed and approved by the NR , or the components have been evaluated in accordance with REPORTA an NRC approved methodology 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. SHUTDOWN MARGIN 1.30 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. MILLSTONE - UNIT 3 1-5 Amendment No. 69, 8+, +88, ~, w

1. 1 Definitions DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE North Anna Units 1 and 2 FOR INFORMATION ONLY Definitions 1.1 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-13lm, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.I of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its 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. 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 and approved by the NRC7,orthe components have been evaluated in aceordance with an NRC approved methodology, LEAKAGE shall be:

a. Identified LEAKAGE
1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (continued) 1.1-3 Amendments ~3,9.

1.1 Definitions OPERABLE-OPERABILITY (continued) PHYSICS TESTS QUADRANT POWER TILT RATIO (QPTR) RATED THERMAL POWER (RTP) REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SOM) North Anna Units 1 and 2 FOR INFORMATION ONLY Definitions 1.1 component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2940 MWt. The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 and approved by the NRC~, or the components have been evaluated in accordance with an NRC approved methodology, SOM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: (continued) 1.1-5 Amendments 258/239

DEFINITIONS 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. 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 the methodology for verification have been previously reviewed and approved by the NRC_1 FREQUENCY NOTATION or the components have been evaluated in accordance wI an NRC aooroved methodoloav. 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. 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 CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c.

Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage). MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay. SUMMER - UNIT 1 1-3 Amendment No. 44e, 4+9

DEFINITIONS PURGE* PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature. pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement QUAOAANT POWER TILT RATIO 1.24 QUAO.~NT POWER TILT RATIO shall be the ratio of tne maximum upper excore detector calibrated *output to the average of the upper' excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2900 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time intesval from when the* monitored parameter exceeds its bip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 the methodology for verification have been previously reviewed and* approved by the NRC , or the components have been evaluated in accordance REPORTABLE EVENT with an NRG approved methodology. 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR 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 withdra1Ml. SLAVE RELAY TEST 1.29 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. 1.30 Not Used SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. SUMMER - UNIT 1 1-5 Amendment No. as, :i041 117, 13~. 14.6 REVISED TECHNICAL SPECIFICATIONS PAGES Dominion Energy Connecticut, Inc. (DENC) Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC) Millstone Power Station Unit 3 North Anna Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microCurie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-13lm, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." 1.12 DELETED ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) 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. 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 the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology. 1.14 DELETED FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1. 1. LEAKAGE 1.16 LEAKAGE shall be: 1.16.l CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
b.

MILLSTONE - UNIT 3 1-3 Amendment No. -84, 7, -H6,..J-8.'.7, ~, ~.~.~

DEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or
c. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

1.16.3 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in a RCS component body, pipe wall, or vessel wall, and 1.16.4 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all LEAKAGE which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. MASTER RELAY TEST 1.17 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include continuity check of each associated slave relay. MEMBER(S) OF THE PUBLIC 1.18 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant. The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location. OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s). MILLSTONE - UNIT 3 1-4 Amendment No.~

DEFINITIONS OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2. PHYSICS TESTS 1.21 PHYS I CS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. 1.22 DELETED PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors sha11 be used for computing the average. RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3650 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.28 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. 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 the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology. MILLSTONE - UNIT 3 1-5 Amendment No. 69, -l-8-1-, 88, ;BS, ~

DEFINITIONS REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. SHUTDOWN MARGIN 1.30 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 inse1ied except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SITE BOUNDARY 1.31 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation. STAGGERED TEST BASIS 1.34 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, and

b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval. THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy. MILLSTONE - UNIT 3 1-6 Amendment No. B-8

DEFINITIONS 1.37 DELETED UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. VENTING 1.39 VENTING shall be the controlled process of discharging ail-or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. 1.40 Deleted I.41 Deleted CORE OPERATING LIMITS REPORT (COLR) 1.42 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Unit Operation within these operating limits is addressed in individual specifications. 1.43 Deleted 1.44 Deleted MILLSTONE - UNIT 3 1-7 Amendment No. :3-9, §G, 6G, :R::, -WG, -l-89-,~,273

NOTATION s D w M Q SA R SIU N.A. p SFCP MILLSTONE - UNIT 3 February 25, 2014 TABLE 1.1 FREQUENCY NOTATION 1-8 FREQUENCY At least once per 12 hours. At least once per 24 hours. At least once per 7 days. At least once per 31 days. At least once per 92 days. At least once per 184 days. At least once per 18 months. Prior to each reactor startup. Not applicable. Completed prior to each release. At the frequency specified in the Surveillance Frequency Control Program Amendment No. 258

1.1 Definitions DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE North Anna Units 1 and 2 Definitions 1.1 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-13lm, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.I of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its 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. 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 and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology. LEAKAGE shall be:

a. Identified LEAKAGE
1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (continued) 1.1-3 Amendments
1. 1 Definitions LEAKAGE (continued)

MASTER RELAY TEST MODE OPERABLE-OPERABILITY North Anna Units 1 and 2 Definitions 1.1

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;
c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, (continued) 1.1-4 Amendments 258/239

1.1 Definitions OPERABLE-OPERABILITY (continued) PHYSICS TESTS QUADRANT POWER TILT RATIO (QPTR) RATED THERMAL POWER (RTP) REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SOM) North Anna Units 1 and 2 Definitions 1.1 component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2940 MWt. The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. 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 and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology. SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: 1.1-5 Amendments

1.1 Definitions SHUTDOWN MARGIN (SOM) (continued) SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT) North Anna Units 1 and 2 Definitions 1.1

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and
b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps. A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps. 1.1-6 Amendments 258/239

DEFINITIONS 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. 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 the methodology for verification have been previously reviewed and approved by the NRC or the components have been evaluated in accordance with an NRG approved methodology. FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. 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 CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c.

Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage). MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay. SUMMER - UNIT 1 1-3 Amendment No. 146,179

DEFINITIONS PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RA TIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2900 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 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. 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 the methodology for verification have been previously reviewed and approved by the NRG, or the components have been evaluated in accordance with an NRC approved methodology. REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50. 73 to 10 CFR 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. SLAVE RELAY TEST 1.29 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. 1.30 Not Used SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. SUMMER - UNIT 1 1-5 Amendment No. 35, 104, 117, 133, 146, PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP) Dominion Energy Connecticut, Inc. (DENC) Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC) Millstone Power Station Unit 3 North Anna Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

~ LBDCR 18 MP3 00~ Mttreh 15, 2018 INSTRUMENTATION BASES 3/4.3.l and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test 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 must be demonstrated by test. Detector response times may be measured by the in-situ online noise analysis-response time degradation method described in the Westinghouse Topical Report, "The Use of Process Noise Measurements to Determine Response Characteristics of Protection Sensors in U.S. Plants," dated August 1983. WCAP-14036, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic 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. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.~ The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: ( 1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) quench spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start, (10) service water pumps start and automatic valves position, and (11) Control Room isolates. MILLSTONE - UNIT 3 B 3/4 3-2b Amendment No. '3-,, +98-, ~

INSERT The response time may be verified for components that replace the components that were previously evaluated in WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996 and WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Revision 2, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing".

BASES SURVEILLANCE REQUIREMENTS FOR INFORMATION ONLY SR 3.3.1.16 (continued) RTS Instrumentation B 3.3.1 time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured. Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" (Ref. 10) 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 must be demonstrated by test. WCAP-14036-P-A Revision 1 "Elimination of Periodic Protection Channel Response Time Tests" (Ref. 11) provides the basis and the methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic 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. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. ~IN_S_E_R_T_A~ (continued) North Anna Units 1 and 2 B 3.3.1-59 Revision 46

BASES SURVEILLANCE REQUIREMENTS REFERENCES FOR INFORMATION ONLY SR 3.3.1.16 (continued) RTS Instrumentation B 3.3.1 SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Response of neutron flux signal portion of the channel time shall be measured from the detector or input of the first electronic component in the channel. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually _ii nstantaneous response.

1. UFSAR, Chapter 7.
2. UFSAR, Chapter 6.
3. UFSAR, Chapter 15.
4. IEEE-279-1971.
5. 10 CFR 50.49.
6. RTS/ESFAS Setpoint Methodology Study Report EE-0116).

(Technical

7. WCAP-10271-P-A, Supplement 1, Rev. 1, June 1990 and WCAP-14333-P-A, Rev. 1, October 1998.
8. Technical Requirements Manual.
9. Regulatory Guide 1.105, Revision 3, "Setpoints for Safety Related Instrumentation."
10. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements,"

January 1996.

11. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.
12. Attachment 1 to TSTF-569, Revision 2, "Methodology to Eliminate 0 ressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing".

North Anna Units 1 and 2 B 3.3.1-60 Revision%

BASES SURVEILLANCE REQUIREMENTS FOR INFORMATION ONLY SR 3.3.2.9 (continued) ESFAS Instrumentation B 3.3.2 For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured. Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements 11 (Ref. 10) 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 must be demonstrated by test. WCAP-14036-P-A Revision 1 11Elimination of Periodic Protection Channel Response Time Tests" (Ref. 11) provides the basis and the methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic 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. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter. -<----I INSERT A I (continued) North Anna Units 1 and 2 B 3.3.2-48 Revision M

BASES SURVEILLANCE REQUIREMENTS REFERENCES FOR INFORMATION ONLY ESFAS Instrumentation B 3.3.2 SR 3.3.2.9 (continued) The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours after reaching 1005 psig in the SGs.

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UFSAR, Chapter 6. UFSAR, Chapter 7. UFSAR, Chapter 15. IEEE-279-1971. 10 CFR 50.49. RTS/ESFAS Setpoint Methodology Study (Technical Report EE-0116). NUREG-1218, April 1988. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990 and WCAP-14333-P-A, Rev. 1, October 1998. Technical Requirements Manual. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.

11. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.
12. Attachment 1 to TSTF-569, Revision 2, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)

Response Time Testing". North Anna Units 1 and 2 B 3.3.2-49 Revision 46

INSERT A The response time may be verified for components that replace the components that were previously evaluated in Ref. 1 O and Ref. 11, provided that the components have been evaluated in accordance with the NRG approved methodology as discussed in Attachment 1 to TSTF-569, Revision 2, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing," (Ref. 12).

INSTRUMENTATION BASES REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continued) WCAP-14036-P-A, Revision 1, "Elimination of Periodic Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component into operational service and re-verified following maintenance or modification that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for the repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing element of a transmitter~ --1 INSERT A I Westinghouse letter CGE-00-018, dated March 28, 2000, provided an evaluation of the Group 05 (11 NLP and 6NSA) 7300 process cards. These cards were revised after the submittal of WCAP-14036, Revision 1. This letter concluded that the failure modes and effects analysis (FMEA) performed for the older versions of these cards and documented in WCAP-14036-P-A, Revision 1, is applicable for these Group 05 cards. The bounding time response values determined by test and evaluation and reported in the WCAP are valid for these redesigned cards. The Engineered Safety Features response times specified in Table 3.3-5 which include sequential operation of the RWST and VCT valves (Notes 2 and 3) are based on values assumed in the non-LOCA safety analyses. These analyses are for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction isolation valves are closed following opening of the RWST charging pumps suction valves. When the sequential operation of the RWST and VCT valves is not included in the response times (Note 1) the values specified are based on the LOCA analyses. The LOCA analyses take credit for injection flow regardless of the source. Verification of the response times specified in Table 3.3-5 will assure that the assumptions used for the LOCA and non-LOCA analyses with respect to the operation of the VCT and RWST valves are valid. The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident 1) safety injection pumps start and automatic valves position, 2) reactor trip, 3) feedwater isolation, 4) startup of the emergency diesel generators, 5) containment spray pumps start and automatic valves position, 6) containment isolation, 7) steam line isolation, 8) turbine trip, 9) auxiliary feedwater pumps start and automatic valves position, 10) containment cooling fans start and automatic valves position, 11) essential service water pumps start and automatic valves position, and 12) control room isolation and ventilation systems start. SUMMER - UNIT 1 B 3/4 3-1c Amendment No. 67, 146, 158,177,209

INSERT A The response time may be verified for components that replace the components that were previously evaluated in WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996 and WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995, provided that the components have been evaluated in accordance with the NRG approved methodology as discussed in Attachment 1 to TSTF-569, Revision 2, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing".}}

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