ML21279A231

Application to Revise Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant Leakage Requirements Using the Consolidated Line-Item Improvement

ML21279A231
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford
Issue date:	10/06/2021
From:	Gaston R Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNRO2021-00020
Download: ML21279A231 (80)

Text

Entergy Operations, Inc 1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 CNRO2021-00020 October 6, 2021 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Application to Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" using the Consolidated Line-Item Improvement Process Arkansas Nuclear One, Unit 1 and 2 River Bend Station, Unit 1 NRC Docket No. 50-313 and 50-368 NRC Docket No. 50-458 Renewed Facility Operating License Renewed Facility Operating License No. DPR-51 and NPF-6 No. NPF-47 Grand Gulf Nuclear Station, Unit 1 Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-416 NRC Docket No. 50-382 Renewed Facility Operating License Renewed Facility Operating License No.

No. NPF-29 NPF-38 Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TS) for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Grand Gulf Nuclear Station, Unit 1 (GGNS), River Bend Station, Unit 1 (RBS), and Waterford Steam Electric Station, Unit 3 (Waterford-3).

Entergy requests adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements,"

which is an approved change to the Standard Technical Specifications (STS), into the ANO-1, ANO-2, GGNS, RBS, and Waterford-3 Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected and adds a Required Action when pressure boundary leakage is identified. The change is requested as part of the Consolidated Line-Item Improvement Process (CLIIP).

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

CNRO2021-00020 Page 2 of 3 Approval of the proposed amendment is requested by May 31, 2022. Once approved, the amendment shall be implemented within 90 days This letter contains no new regulatory commitments.

Should you have any questions or require additional information, please contact Phil Couture, Sr. Manager, Fleet Regulatory Assurance, at 601-368-5102.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), a copy of this application, with attachments, is being provided to the designated State Officials.

I declare under penalty of perjury the foregoing is true and correct. Executed on October 6, 2021.

Respectfully, Ron Gaston RWG/chm

Enclosure:

Description and Assessment of the Proposed Change Attachments to

Enclosure:

1. Technical Specification Page Markups

2. Technical Specification Bases Page Markups (Information Only)

3. Retyped Technical Specification Pages

CNRO2021-00020 Page 3 of 3 cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - ANO NRC Senior Resident Inspector - GGNS NRC Senior Resident Inspector - RBS NRC Senior Resident Inspector - Waterford-3 NRC Project Manager - Entergy Fleet NRC Project Manager - ANO NRC Project Manager - GGNS NRC Project Manager - RBS NRC Project Manager - Waterford-3 Designated State Official - Arkansas Louisiana Department of Environmental Quality NRC Project Manager State Health Officer, Mississippi State Department of Health

Enclosure CNRO2021-00020 Description and Assessment of the Proposed Change

CNRO2021-00020 Enclosure Page 1 of 7 TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION ............................................................................................... 2 2.0 ASSESSMENT ................................................................................................................... 2 2.1 Applicability of Safety Evaluation .................................................................................... 2 2.2 Variations ........................................................................................................................ 2 2.2.1 ANO-1 ..................................................................................................................... 2 2.2.2 ANO-2 ..................................................................................................................... 2 2.2.3 GGNS and RBS ...................................................................................................... 3 2.2.4 Waterford-3 ............................................................................................................. 4

3.0 REGULATORY ANALYSIS

................................................................................................ 5 3.1 No Significant Hazards Consideration Analysis.............................................................. 5 3.2 Conclusions .................................................................................................................... 6 4.0 ENVIRONMENTAL EVALUATION .................................................................................... 6 5.0 ATTACHMENTS ................................................................................................................ 7

CNRO2021-00020 Enclosure Page 2 of 7 DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION Entergy Operations, Inc.(Entergy) requests adoption of Technical Specification Task Force (TSTF) Traveler TSTF-554, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the Arkansas Nuclear One (ANO), Units 1 and 2 (ANO-1 and ANO-2), Grand Gulf Nuclear Station, Unit 1 (GGNS),

River Bend Station, Unit 1 (RBS), and Waterford Steam Electric Station, Unit 3 (Waterford-3)

Technical Specifications (TS) using the Consolidated Line Item Improvement Process. The proposed amendment revises the TS definition of "Leakage" and the Reactor Coolant System (RCS) Operational Leakage TS to clarify the requirements.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation Entergy has reviewed the NRC safety evaluation (SE) for TSTF-554 provided to the TSTF by letter dated December 18, 2020. This review included a review of the NRC's SE, as well as the information provided in TSTF-554. As described herein, Entergy has concluded that the justifications presented in TSTF-554 and the SE prepared by the NRC are applicable to ANO-1, ANO-2, GGNS, RBS, and Waterford-3 and justify this amendment for the incorporation of the changes to the ANO-1, ANO-2, GGNS, RBS, and Waterford-3 TS.

2.2 Variations Entergy is proposing the following variations from the TS changes described in TSTF-554 or the applicable parts of the NRC staffs SE. The ANO-2, GGNS, RBS, and Waterford-3 TS contain requirements that differ from the Standard Technical Specifications (STS) on which TSTF-554 was based but are encompassed in the TSTF-554 justification.

2.2.1 ANO-1 There are no variations noted for ANO-1.

2.2.2 ANO-2 The ANO-2 TS contain requirements that differ from the STS on which TSTF-554 was based but are encompassed in the TSTF-554 justification.

• The ANO-2 TS are based on NUREG-0212, "Standard Technical Specifications, Combustion Engineering Plants." TSTF-554 is based on NUREG-1432 of the same title.

• In the STS on which TSTF-554 is based, all leakage definitions appear as part of a single defined term, "LEAKAGE." In the ANO-2 TS, the terms related to leakage are separate definitions. However, the changes to the affected defined terms "IDENTIFIED LEAKAGE" and "PRESSURE BOUNDARY LEAKAGE" are the same as the changes made to those terms in TSTF-554.

CNRO2021-00020 Enclosure Page 3 of 7

• In the ANO-2 TS, the "Operational Leakage" specification is numbered 3.4.6.2. In the STS on which TSTF-554 is based, the TS is numbered 3.4.13.

• In TSTF-554, a new Action A is created that is applicable when there is pressure boundary leakage. In the ANO-2 TS, existing Action a applies when there is pressure boundary leakage or primary to secondary leakage not within limit. The ANO-2 TS are revised such that Action a is applicable when primary to secondary leakage is not within limit and a new Action b is created that applies to the existence of pressure boundary leakage. The existing ANO-2 Actions b and c are renamed Actions c and d.

• In the ANO-2 TS, shutdown requirements are included in each Action. The TSTF-554 shutdown actions in STS Condition C are incorporated into new ANO-2 Action b. The requirements are consistent with TSTF-554.

• The ANO-2 TS Bases for the "Operational Leakage" specification differ from the STS Bases. However, the applicable TSTF-554 Bases changes are incorporated into the ANO-2 TS Bases.

2.2.3 GGNS and RBS The current GGNS and RBS TS 3.4.5, "RCS Operational Leakage," Action B differs slightly from the same Action in Revision 4 of NUREG-1434, "Standard Technical Specifications, General Electric Plants, BWR/6" (i.e., the basis for TSTF-554). Specifically, the NUREG-1434 TS 3.4.5 Action B includes Required Actions B.1 and B.2, which are linked with an "OR" connector, each with a 4-hour Completion Time (CT):

B.1 Reduced LEAKAGE to within limit.

OR B.2 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

Failure to comply with either of the two Required Actions within the specified CT would require the unit to be placed in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

In contrast to NUREG-1434 Action B, the GGNS and RBS TSs only include NUREG-1434 Required Action B.2 as a specified action. This difference does not impact the overall intent of the Action. As stated in the NUREG-1434 TS Bases, an unidentified increase of greater than 2 gallons per minute in leakage within a 4-hour period is an indication of a potential flaw in the Reactor Coolant Pressure Boundary and must be quickly evaluated. Although the increase may not necessarily violate the absolute unidentified leakage limit, certain susceptible components must be determined not to be the source of the leakage increase within the required CT. Thus, current Action B in the GGNS and RBS TSs is consistent with the intent of the NUREG-1434 Action B. In addition, this difference does not affect the applicability of TSTF-554 to the GGNS and RBS TS.

CNRO2021-00020 Enclosure Page 4 of 7 For the GGNS TS 1.1 "Definitions," page 1.0-4, and TS 3.4.5 "RCS Operational LEAKAGE,"

Entergy is making editorial adjustment to the font and margins on the page. No additional content changes are being made other than the text shown as changing for TSTF-554.

2.2.4 Waterford-3 The Waterford-3 TS contain requirements that differ from the STS on which TSTF-554 was based but are encompassed in the TSTF-554 justification.

The Waterford-3 TS are based on NUREG-0212, "Standard Technical Specifications, Combustion Engineering Plants." TSTF-554 is based on NUREG-1432 of the same title.

In the STS on which TSTF-554 is based, all leakage definitions appear as part of a single defined term, "LEAKAGE." In the Waterford-3 TS, the terms related to leakage are separate definitions. However, the changes to the affected defined terms "identified leakage" and "pressure boundary leakage" are the same as the changes made to those terms in TSTF-554.

In the Waterford-3 TS, the "Operational Leakage" specification is numbered 3.4.5.2. In the STS on which TSTF-554 is based, the TS is numbered 3.4.13.

In TSTF-554, a new Action A is created that is applicable when there is pressure boundary leakage. In the Waterford-3 TS, existing Action a applies when there is pressure boundary leakage or primary to secondary leakage not within limit. The Waterford-3 TS are revised such that Action a is applicable when primary to secondary leakage is not within limit and a new Action b is created that applies to the existence of pressure boundary leakage. The existing Waterford-3 Actions b and c are renamed Actions c and d.

In the Waterford-3 TS, shutdown requirements are included in each Action. The TSTF-554 shutdown actions in STS Condition C are incorporated into new Waterford Action b. The requirements are consistent with TSTF-554.

The Waterford-3 TS Bases for the "Operational Leakage" specification differ from the STS Bases. However, the applicable TSTF-554 Bases changes are incorporated into the Waterford TS Bases.

An editorial change that is unrelated to this TSTF is being made to DEFINITION 1.23 "PURGE - PURGING." A preexisting typographical error was identified at the end of the second line. The word "concentration" had been mis-typed as "concentra-." Since we are working on this page of TS, the typographical error is being corrected to "concentration." Correcting this typographical error does not negatively affect the definition.

CNRO2021-00020 Enclosure Page 5 of 7

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis In accordance with Title 10 of the Code of Federal Regulations (CFR) Part 50, Section 50.90, "Application for amendment of license, construction permit, or early site permit,"

Entergy Operations, Inc. (Entergy) requests to adopt Technical Specification Task Force (TSTF) Traveler TSTF-554, "Revise Reactor Coolant Leakage Requirements" into Renewed Facility Operating License, Appendix A, "Technical Specifications" (TS) for Arkansas Nuclear One (ANO) Units 1 and 2 (ANO-1 and ANO-2), Grand Gulf Nuclear Station, Unit 1 (GGNS), River Bend Station, Unit 1 (RBS), and Waterford Steam Electric Station, Unit 3 (Waterford-3) using the Consolidated Line Item Improvement Process.

The proposed amendment revises TS definitions of "leakage," clarifies the requirements when pressure boundary leakage is detected and adds an Action when pressure boundary leakage is identified.

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

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

Response: No The proposed change revises the TS definitions of "leakage," clarifies the requirements when pressure boundary leakage is detected and adds an Action when pressure boundary leakage is identified.

The proposed change revises the definition of pressure boundary leakage.

Pressure boundary leakage is a precursor to some accidents previously evaluated.

The proposed change expands the definition of pressure boundary leakage by eliminating the qualification that pressure boundary leakage must be from a "nonisolable" flaw. A new TS Action is created which requires isolation of the pressure boundary flaw from the Reactor Coolant System. This new action provides assurance that the flaw will not result in any accident previously evaluated.

Pressure boundary leakage, and the actions taken when pressure boundary leakage is detected, is not assumed in the mitigation of any accident previously evaluated. As a result, the consequences of any accident previously evaluated are unaffected.

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

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

Response: No

CNRO2021-00020 Enclosure Page 6 of 7 The proposed change revises the TS definitions of "leakage," clarifies the requirements when pressure boundary leakage is detected and adds an Action when pressure boundary leakage is identified. The proposed change does not alter the design function or operation of the RCS. The proposed change does not alter the ability of the RCS to perform its design function. Since pressure boundary leakage is an evaluated accident, the proposed change does not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

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 change involve a significant reduction in a margin of safety?

Response: No The proposed change revises the TS definitions of "leakage," clarifies the requirements when pressure boundary leakage is detected and adds an Action when pressure boundary leakage is identified. The proposed change does not affect the initial assumptions, margins, or controlling values used in any accident analysis. The amount of leakage allowed from the RCS is not increased. The proposed change does not affect any design basis or safety limit or any Limiting Condition for Operation.

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

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

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

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 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).

CNRO2021-00020 Enclosure Page 7 of 7 Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

5.0 ATTACHMENTS

1. Technical Specification Page Markups

2. Technical Specification Bases Page Markups (Information Only)

3. Retyped Technical Specification Pages

Enclosure, Attachment 1 CNRO2021-00020 Technical Specification Page Markups ANO-1 ANO-2 GGNS RBS Waterford-3 Number of Plant Affected TS Pages Arkansas Nuclear One - Unit 1 4 Arkansas Nuclear One - Unit 2 3 Grand Gulf Nuclear Station, Unit 1 3 River Bend Station, Unit 1 3 Waterford Steam Electric Station, Unit 3 4 Total Number of Pages in Attachment 1 17

Enclosure, Attachment 1 CNRO2021-00020 Technical Specification Page Markups Arkansas Nuclear One - Unit 1 Technical Specification Affected Number TS Number Title of Pages 1.1 Definition - LEAKAGE 2 3.4.13 RCS Operational LEAKAGE 2 Total ANO-1 TS Pages 4

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion."

DOSE EQUIVALENT XE-133 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-131m, 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."

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known toeither 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).;

ANO-1 1.1-3 Amendment No. 215,224,243,257,270,

Definitions 1.1 1.1 Definition (continued)

LEAKAGE (continued) b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and 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. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MODE 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.

OPERABLE-OPERABILITY 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, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS 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 the SAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

Power in any Core Quadrant QPT = 100 1 Average Power in all Quadrants ANO-1 1.1-4 Amendment No. 215,243,270,

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

d. 150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary A.1 Isolate affected component, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE exists. pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

BA. RCS unidentified or BA.1 Reduce LEAKAGE to within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> identified LEAKAGE not limits.

within limits, except for primary to secondary LEAKAGE.

ANO-1 3.4.13-1 Amendment No. 215,224,

RCS Operational LEAKAGE 3.4.13 CB. Required Action and CB.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

CB.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

ANO-1 3.4.13-xx Amendment No. 215,224,

Enclosure, Attachment 1 CNRO2021-00020 Technical Specification Page Markups Arkansas Nuclear One - Unit 2 Technical Specification Affected Number TS Number Title of Pages 1.14 Definitions - IDENTIFIED LEAKAGE 1 1.16 Definitions - PRESSURE BOUNDARY LEAKAGE 1 3.4.6.2 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 1 Total ANO-2 TS Pages 3

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c. Digital computer channels - The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SHUTDOWN MARGIN 1.13 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 control element assemblies are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 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 to 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).

ARKANSAS - UNIT 2 1-3 Amendment No. 157,220,255,266, 319,322,

DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

AZIMUTHAL POWER TILT - Tq 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT I-131 1.18 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

DOSE EQUIVALENT XE-133 1.19 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-131m, 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.

1.20 Deleted FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

ARKANSAS - UNIT 2 1-4 Amendment No. 157,255,266,282, 293,315,

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE or any primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

cb. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and primary to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

dc. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two valves* in each high pressure line having a non-functional valve and 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• These valves may include check valves for which the leakage rate has been verified, manual valves or automatic valves. Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.

ARKANSAS - UNIT 2 3/4 4-14 Amendment No. 184,266,280, Order dated 4/20/81

Enclosure, Attachment 1 CNRO2021-00020 Technical Specification Page Markups Grand Gulf Nuclear Station, Unit 1 Technical Specification Affected Number TS Number Title of Pages 1.1 Definitions - LEAKAGE 1 3.4.5 RCS Operational LEAKAGE 2 Total GGNS TS Pages 3

Definitions 1.1 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may (continued)

GRAND GULF 1.0-4 Amendment No. 120,

RCS Operational LEAKAGE 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Operational LEAKAGE LCO 3.4.5 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 5 gpm unidentified LEAKAGE;

c. 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d. 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary A.1 Isolate affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE exists. component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

BA. Unidentified LEAKAGE BA.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> not within limit. within limits.

OR Total LEAKAGE not within limit.

GRAND GULF 3.4-12 Amendment No. 120,

RCS Operational LEAKAGE 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME CB. Unidentified LEAKAGE CB.1 Verify source of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> increase not within unidentified LEAKAGE limit. increase is not service sensitive type 304 or type 316 austenitic stainless steel.

DC. Required Action and DC.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A or AND B not met.

DC.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY In accordance with SR 3.4.5.1 Verify RCS unidentified LEAKAGE, total LEAKAGE, the Surveillance and unidentified LEAKAGE increase are within Frequency Control limits.

Program GRAND GULF 3.4-13 Amendment No. 120, 219,

Enclosure, Attachment 1 CNRO2021-00020 Technical Specification Page Markups River Bend Station, Unit 1 Technical Specification Affected Number TS Number Title of Pages 1.1 Definitions - LEAKAGE 1 3.4.5 RCS Operational LEAKAGE 2 Total RBS TS Pages 3

Definitions 1.1 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

(continued)

RIVER BEND 1.0-4 Amendment No. 81,

RCS Operational LEAKAGE 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Operational LEAKAGE LCO 3.4.5 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 5 gpm unidentified LEAKAGE;

c. 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d. 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary A.1 Isolate affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE exists. component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

BA. Unidentified LEAKAGE not BA.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within limit. within limits.

OR Total LEAKAGE not within limit.

RIVER BEND 3.4-12 Amendment No. 81, 196,

RCS Operational LEAKAGE 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME CB. Unidentified LEAKAGE CB.1 Verify source of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> increase not within limit. unidentified LEAKAGE increase is not service sensitive type 304, type 316 austenitic stainless steel, or other intergranular stress corrosion cracking susceptible material.

DC. Required Action and DC.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A or B AND not met.

DC.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify RCS unidentified LEAKAGE, total LEAKAGE, In accordance with and unidentified LEAKAGE increase are within limits. the Surveillance Frequency Control Program RIVER BEND 3.4-13 Amendment No. 81, 196,

Enclosure, Attachment 1 CNRO2021-00020 Technical Specification Page Markups Waterford Steam Electric Station, Unit 3 Technical Specification Affected Number TS Number Title of Pages 1.14 DEFINITIONS - IDENTIFIED LEAKAGE 1 1.21 & DEFINITIONS - PRESSURE BOUNDARY LEAKAGE 1 1.23 DEFINITIONS - PURGE - PURGING 3.4.5.2 OPERATIONAL LEAKAGE 2 Total Waterford-3 TS Pages 4

DEFINITIONS IDENTIFIED LEAKAGE (Continued)

b. Leakage into the containment atmosphere from sources that are both specifically located and known to 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).

MEMBER(S) OF THE PUBLIC 1.15 MEMBER(S) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.16 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specification 6.9.1.7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.17 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).

OPERATIONAL MODE - MODE 1.18 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.

WATERFORD - UNIT 3 1-4 AMENDMENT NO. 68, 84, 116, 204,

DEFINITIONS PHYSICS TEST 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (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.

PLANAR RADIAL PEAKING FACTOR - Fxy 1.20 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non isolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that process-ing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentra-tion or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

Editorial Correction of typographical error - "concentra-" is being corrected to "concentration" WATERFORD - UNIT 3 1-5 AMENDMENT NO. 68, 204,

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 75 gallons per day primary to secondary leakage, through any one steam generator (SG),

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 1 gpm leakage at a Reactor Coolant System pressure of 2250 +/- 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

cb. With any Reactor Coolant System operational leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE, primary to secondary leakage, and leakage from Reactor Coolant System pressure isolation valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

dc. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

WATERFORD - UNIT 3 3/4 4-18 AMENDMENT NO. 197,199, 204, 249,

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS NOTE: Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

4.4.5.2.1 Reactor Coolant System leakages, except for primary to secondary leakage, shall be demonstrated to be within each of the above limits by performance of a Reactor Coolant System water inventory balance in accordance with the Surveillance Frequency Control Program.

4.4.5.2.2 Primary to secondary leakage shall be verified to be 75 gallons per day through any one SG in accordance with the Surveillance Frequency Control Program.

SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.2.3 Each Reactor Coolant System pressure isolation valve specified in Table 3.4-1, Section A and Section B, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. In accordance with the Surveillance Frequency Control Program,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve,

d. Following valve actuation for valves in Section B due to automatic or manual action or flow through the valve:

1. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying valve closure, and

2. Within 31 days by verifying leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.5.2.4 Each Reactor Coolant System pressure isolation valve power-operated valve specified in Table 3.4-1, Section C, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. In accordance with the Surveillance Frequency Control Program, and

b. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

WATERFORD - UNIT 3 3/4 4-19 AMENDMENT NO. 96,107, 204, 249,

Enclosure, Attachment 2 CNRO2021-00020 Technical Specification Bases Page Markups (Information Only)

ANO-1 ANO-2 GGNS RBS Waterford-3 Number of Plant Affected TS Bases Pages Arkansas Nuclear One - Unit 1 4 Arkansas Nuclear One - Unit 2 2 Grand Gulf Nuclear Station, Unit 2 4 River Bend Station, Unit 1 4 Waterford Steam Electric Station, Unit 3 4 Total Number of Pages in Attachment 2 18

Enclosure, Attachment 2 CNRO2021-00020 Technical Specification Bases Page Markups (Information Only)

Arkansas Nuclear One - Unit 1 Technical Specification Bases Affected TS Bases Number Number Title of Pages B 3.4.13 RCS Operational LEAKAGE 4 Total ANO-1 TS Bases Pages 4

RCS Operational LEAKAGE B 3.4.13 APPLICABLE SAFETY ANALYSES (continued)

Primary to secondary LEAKAGE is a factor in the radioactivity releases resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The SAR (Ref. 4) analysis for SGTR assumes the contaminated secondary fluid is released via turbine bypass valves to the condenser and through the MSSVs and through the ADVs to the atmosphere. The 150 gpd primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential compared to the tube rupture leakage.

The safety analysis for the SLB accident assumes 1 gpm total primary to secondary LEAKAGE as an initial condition. The dose consequences resulting from the MSLB accident are a small fraction of 10 CFR 50.67 limits.

RCS leakage detection capabilities and methods are identified and discussed in SAR Section 4.2.3.8 (Ref. 5) and in the Bases for LCO 3.4.15, "RCS Leakage Detection Instrumentation."

In MODES 1 and 2, RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36 (Ref. 6).

In MODES 3 and 4, RCS operational LEAKAGE satisfies Criterion 4 of 10 CFR 50.36.

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pPressure boundary LEAKAGE is prohibitedallowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further RCPB deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the reactor building air monitoring and reactor building sump level monitoring equipment can detect within a reasonable time period.

Separating the sources of LEAKAGE (i.e., LEAKAGE from an identified source versus LEAKAGE from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary. Reactor coolant pump (RCP) controlled bleedoff is a normal function and is not considered as LEAKAGE.

ANO-1 B 3.4.13-2 Amendment No. 215 Rev. 7,33,45,

RCS Operational LEAKAGE B 3.4.13 LCO (continued)

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS makeup system. Identified LEAKAGE includes LEAKAGE to the reactor building from specifically known and located sources and LEAKAGE through a SG to the secondary system, but does not include pressure boundary LEAKAGE or RCP controlled bleedoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 7). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures and to assure the safety analysis is bounding.

APPLICABILITY In MODES 1, 2, 3, and 4, the LEAKAGE limits are required because the RCS is pressurized and the potential for RCPB LEAKAGE is greatest.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potential for LEAKAGE.

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through RCS pressure isolation valves (PIVs) and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leaktight. If both valves in series leak and result in a loss of coolant mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS A.1 If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location.

ANO-1 B 3.4.13-3 Amendment No. 215 Rev. 7,45,

RCS Operational LEAKAGE B 3.4.13 ACTIONS (continued)

A.1 (continued)

Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

BA.1 If unidentified LEAKAGE, or identified LEAKAGE, or both, are in excess of the LCO limits, the LEAKAGE must be reduced to within limits within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

CB.1 and CB.2 If any pressure boundary LEAKAGE exists, or if primary to secondary LEAKAGE is not within limit, or any of the Required Actions and associated Completion Times cannot be metif the Required Action and associated Completion Time of Condition A is not met, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to 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 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The Completion Times allowed are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, the pressure stresses acting on the RCPB are much lower and further deterioration is much less likely.

SURVEILLANCE REQUIREMENTS SR 3.4.13.1 Verifying RCS LEAKAGE within the LCO limits ensures that the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and may be positively identified by inspection. Total LEAKAGE is determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows) at or near operating pressure. The Surveillance is modified by two notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation at or near operating pressure (i.e., at or near 2155 psig). The 12-hour allowance provides sufficient time to collect and process all necessary data after stable unit conditions are established.

ANO-1 B 3.4.13-4 Amendment No. 215 Rev. 7,67,

RCS Operational LEAKAGE B 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.13.1 (continued)

Steady state operation is required to perform a proper water inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP pump seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the reactor building atmosphere radioactivity and the reactor building sump level. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation.

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.16, Steam Generator Tube Integrity, should be evaluated.

The 150 gallons per day limit is measured at room temperature as described in Reference 8.

The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with EPRI Guidelines (Ref. 8).

ANO-1 B 3.4.13-5 Amendment No. 215 Rev. 7,67,

Enclosure, Attachment 2 CNRO2021-00020 Technical Specification Bases Page Markups (Information Only)

Arkansas Nuclear One - Unit 2 Technical Specification Bases Affected TS Bases Number Number Title of Pages B 3/4.4.6.2 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 2 Total ANO-2 TS Bases Pages 2

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

The limit of 150 gallons per day through any one SG is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines which states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

The 150 gallons per day limit is measured at room temperature as described in EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines. The operational leakage rate limit applies to leakage through any one SG. If it is not practical to assign the leakage to an individual SG, all the primary to secondary leakage should be conservatively assumed to be from one SG.

For primary to secondary leakage determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. The surveillance frequency is a reasonable interval to trend primary to secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI Pressurized Water Reactor Primary-to-Secondary Leak Guidelines.

PRESSURE BOUNDARY LEAKAGE of any magnitude is prohibited as the leak itself could cause further reactor coolant pressure boundary (RCPB) deterioration, resulting in higher leakage unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. If PRESSURE BOUNDARY LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation ARKANSAS - UNIT 2 B 3/4 4-9 Amendment No. 92,124,242 Rev. 4,19,34,39,47,56,63,73,

boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal leakage past the isolation device is acceptable as it will limit RCS leakage and is included in IDENTIFIED LEAKAGE or UNIDENTIFIED LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

ARKANSAS - UNIT 2 B 3/4 4-9 Amendment No. 92,124,242 Rev. 4,19,34,39,47,56,63,73,

Enclosure, Attachment 2 CNRO2021-00020 Technical Specification Bases Page Markups (Information Only)

Grand Gulf Nuclear Station, Unit 1 Technical Specification Bases Affected TS Bases Number Number Title of Pages B 3.4.5 RCS Operational LEAKAGE 4 Total GGNS TS Bases Pages 4

RCS Operational LEAKAGE B 3.4.5 BASES (continued)

APPLICABLE The allowable RCS operational LEAKAGE limits are based on SAFETY ANALYSES the predicted and experimentally observed behavior of pipe cracks. The normally expected background LEAKAGE due to equipment design and the detection capability of the instrumentation for determining system LEAKAGE were also considered. The evidence from experiments suggests, for LEAKAGE even greater than the specified unidentified LEAKAGE limits, the probability is small that the imperfection or crack associated with such LEAKAGE would grow rapidly.

The unidentified LEAKAGE flow limit allows time for corrective action before the RCPB could be significantly compromised. The 5 gpm limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Crack behavior from experimental programs (Refs. 4 and 5) shows leak rates of hundreds of gallons per minute will precede crack instability (Ref. 6).

The low limit on increase in unidentified LEAKAGE assumes a failure mechanism of intergranular stress corrosion cracking (IGSCC) that produces tight cracks. This flow increase limit is capable of providing an early warning of such deterioration.

No applicable safety analysis assumes the total LEAKAGE limit. The total LEAKAGE limit considers RCS inventory makeup capability and drywell floor sump capacity.

RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pPressure boundary LEAKAGE is prohibited allowed, being indicative of material degradation. LEAKAGE of this type is unacceptable as the leak itself could cause further RCPB deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

(continued)

GRAND GULF B 3.4-23 Revision No. 0

RCS Operational LEAKAGE B 3.4.5 BASES LCO b. Unidentified LEAKAGE (continued)

Five gpm of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the drywell atmospheric monitoring, drywell sump level monitoring, and drywell air cooler condensate flow rate monitoring equipment can detect within a reasonable time period. Separating the sources of leakage (i.e.,

leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.Violation of this LCO could result in continued degradation of the RCPB.

c. Total LEAKAGE The total LEAKAGE limit is based on a reasonable minimum detectable amount. The limit also accounts for LEAKAGE from known sources (identified LEAKAGE). Violation of this LCO indicates an unexpected amount of LEAKAGE and, therefore, could indicate new or additional degradation in an RCPB component or system.

d. Unidentified LEAKAGE Increase An unidentified LEAKAGE increase of > 2 gpm within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period indicates a potential flaw in the RCPB and must be quickly evaluated to determine the source and extent of the LEAKAGE. The increase is measured relative to the steady state value; temporary changes in LEAKAGE rate as a result of transient conditions (e.g., startup) are not considered. As such, the 2 gpm increase limit is only applicable in MODE 1 when operating pressures and temperatures are established. Violation of this LCO could result in continued degradation of the RCPB.

APPLICABILITY In MODES 1, 2, and 3, the RCS operational LEAKAGE LCO applies because the potential for RCPB LEAKAGE is greatest when the reactor is pressurized.

In MODES 4 and 5, RCS operational LEAKAGE limits are not required since the reactor is not pressurized and stresses in the RCPB materials and potential for LEAKAGE are reduced.

(continued)

GRAND GULF B 3.4-24 Revision No. 0

RCS Operational LEAKAGE B 3.4.5 BASES (continued)

ACTIONS A.1 If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

B.1 With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leakage. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE.

However, the total LEAKAGE limit would remain unchanged.

CB.1 An unidentified LEAKAGE increase of > 2 gpm within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is an indication of a potential flaw in the RCPB and must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to be the source of the LEAKAGE increase within the required Completion Time. For an unidentified LEAKAGE increase greater than required limits, an alternative to reducing LEAKAGE increase to within limits (i.e., reducing the leakage rate such that the current rate is less than the 2 gpm increase in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; either by isolating the source or other possible methods) is to evaluate RCS type 304 and type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased LEAKAGE. This type of piping is very susceptible to IGSCC.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is needed to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down.

(continued)

GRAND GULF B 3.4-25 Revision No. 0

RCS Operational LEAKAGE B 3.4.5 BASES ACTIONS DC.1 and DC.2 If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.7, "RCS Leakage Detection Instrumentation." Sump level is typically monitored to determine actual LEAKAGE rates. However, any method may be used to quantify LEAKAGE within the guidelines of Reference 7.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. 10 CFR 50, Appendix A, GDC 55.

4. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through - Wall Flaws," April 1968.

5. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants,"

October 1975.

6. UFSAR, Section 5.2.5.5.3.

7. Regulatory Guide 1.45, May 1973 with exceptions per UFSAR Appendix 3A.

GRAND GULF B 3.4-26 Revision No. LBDCR 18127

Enclosure, Attachment 2 CNRO2021-00020 Technical Specification Bases Page Markups (Information Only)

River Bend Station, Unit 1 Technical Specification Bases Affected TS Bases Number Number Title of Pages B 3.4.5 RCS Operational LEAKAGE 4 Total RBS TS Bases Pages 4

RCS Operational LEAKAGE B 3.4.5 BASES (continued)

APPLICABLE The allowable RCS operational LEAKAGE limits are based on the SAFETY ANALYSES predicted and experimentally observed behavior of pipe cracks. The normally expected background LEAKAGE due to equipment design and the detection capability of the instrumentation for determining system LEAKAGE were also considered. The evidence from experiments suggests, for LEAKAGE even greater than the specified unidentified LEAKAGE limits, the probability is small that the imperfection or crack associated with such LEAKAGE would grow rapidly.

The unidentified LEAKAGE flow limit allows time for corrective action before the RCPB could be significantly compromised. The 5 gpm limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Crack behavior from experimental programs (Refs. 4 and

5) shows leak rates of hundreds of gallons per minute will precede crack instability (Ref. 6).

The low limit on increase in unidentified LEAKAGE assumes a failure mechanism of intergranular stress corrosion cracking (IGSCC) that produces tight cracks. This flow increase limit is capable of providing an early warning of such deterioration.

No applicable safety analysis assumes the total LEAKAGE limit. The total LEAKAGE limit considers RCS inventory makeup capability and drywell floor sump capacity.

RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pPressure boundary LEAKAGE is allowedprohibited, being indicative of material degradation. LEAKAGE of this type is unacceptable as the leak itself could cause further RCBP deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

(continued)

RIVER BEND B 3.4-23 Revision No. 0,

RCS Operational LEAKAGE B 3.4.5 BASES LCO b. Unidentified LEAKAGE (continued)

Five gpm of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the drywell atmospheric monitoring, drywell sump level monitoring, and drywell air cooler condensate flow rate monitoring equipment can detect within a reasonable time period. Separating the sources of leakage (i.e.,

leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.Violation of this LCO could result in continued degradation of the RCPB.

c. Total LEAKAGE The total LEAKAGE limit is based on a reasonable minimum detectable amount. The limit also accounts for LEAKAGE from known sources (identified LEAKAGE). Violation of this LCO indicates an unexpected amount of LEAKAGE and, therefore, could indicate new or additional degradation in an RCPB component or system.

d. Unidentified LEAKAGE Increase An unidentified LEAKAGE increase of > 2 gpm within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period indicates a potential flaw in the RCPB and must be quickly evaluated to determine the source and extent of the LEAKAGE. The increase is measured relative to the steady state value; temporary changes in LEAKAGE rate as a result of transient conditions (e.g., startup) are not considered. As such, the 2 gpm increase limit is only applicable in MODE 1 when operating pressures and temperatures are established. Violation of this LCO could result in continued degradation of the RCPB.

APPLICABILITY In MODES 1, 2, and 3, the RCS operational LEAKAGE LCO applies because the potential for RCPB LEAKAGE is greatest when the reactor is pressurized.

In MODES 4 and 5, RCS operational LEAKAGE limits are not required since the reactor is not pressurized and stresses in the RCPB materials and potential for LEAKAGE are reduced.

(continued)

RIVER BEND B 3.4-24 Revision No. 0,

RCS Operational LEAKAGE B 3.4.5 BASES (continued)

ACTIONS A.1 If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

B.1 With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the LEAKAGE. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE. However, the total LEAKAGE limit would remain unchanged.

CB.1 An unidentified LEAKAGE increase of > 2 gpm within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is an indication of a potential flaw in the RCPB and must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to be the source of the LEAKAGE increase within the required Completion Time. For an unidentified LEAKAGE increase greater than required limits, an alternative to reducing LEAKAGE increase to within limits (i.e., reducing the leakage rate such that the current rate is less than the 2 gpm increase in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; either by isolating the source or other possible methods) is to evaluate RCS type 304, type 316 austenitic stainless steel piping and other intergranular stress corrosion cracking susceptible material that is subject to high stress or that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased LEAKAGE.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is needed to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down.

C.1 and C.2 RIVER BEND B 3.4-25 Revision No. 0,

RCS Operational LEAKAGE B 3.4.5 BASES ACTIONS DC.1 and DC.2 (continued)

If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.7, "RCS Leakage Detection Instrumentation." Sump level is typically monitored to determine actual LEAKAGE rates. However, any method may be used to quantify LEAKAGE within the guidelines of Reference 7.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. 10 CFR 50, Appendix A, GDC 55.

4. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.

5. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants,"

October 1975.

6. USAR, Section 5.2.5.5.3.

7. Regulatory Guide 1.45, May 1973.

RIVER BEND B 3.4-26 Amendment No. 196,

Enclosure, Attachment 2 CNRO2021-00020 Technical Specification Bases Page Markups (Information Only)

Waterford Steam Electric Station, Unit 3 Technical Specification Bases Affected TS Bases Number Number Title of Pages B 3/4.4.5.2 OPERATIONAL LEAKAGE 4 Total Waterford-3 TS Bases Pages 4

> (DRN 04-1223, Ch. 33)

REACTOR COOLANT SYSTEM BASES (continued)

Monitoring Containment Sump In-Leakage Flow During automatic operation of the containment sump pumps (after a containment sump pump has operated), the flow calculation performed by the plant monitoring computer based on a level change will no longer be accurate since the level in the sump will be lowering. A 20 minute time period has been conservatively determined based on engineering calculations for this equipment operation. In addition, upon reboot of the plant monitoring computer, a period of 10 minutes is required for the leak rate calculation to become available. It has been determined these time periods (independent or combined) of calculation sump in-leakage flow inaccuracies, the instrumentation remains adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour; therefore, the containment sump level instrumentation and the corresponding flow calculation is considered to remain operable.

References

3. 10 CFR 50, Appendix A, Section IV, GDC 30.

4. Regulatory Guide 1.45, Revision 0, dated May 1973.

5. UFSAR, Sections 5.2.5 and 12.3.

< (DRN 04-1223, Ch. 33) 3/4.4.5.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

> (EC-3173 Ch. 53)

For reactor coolant system operational leakage determination, steady state operation is required to perform a proper water balance since calculations during maneuvering are not useful and cannot ensure an accurate measurement is obtained (e.g. when operating in the shutdown cooling mode). The RCS water inventory balance must be performed with the reactor at stable operating pressure and steady state conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RP seal injection and return flows).

Therefore, a Note is added allowing that this surveillance is required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at stable operating pressure after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

WATERFORD - UNIT 3 B 3/4 4-4d CHANGE NO. 33, 38, 52, 53,

REACTOR COOLANT SYSTEM BASES (continued) 3/4.4.5.2 OPERATIONAL LEAKAGE

> (EC-3173 Ch. 53)

The SR 4.4.5.2.1 performance after establishing steady state operation is consistent with the use and application guidance of section 1.4, Frequency, of NUREG-1432 Rev. 3.0, "Improved Standard Technical Specification Combustion Engineering Plants," March 31, 2004.

In accordance with this guidance, the NOTE modifies the required performance of the Surveillance and it is construed to be part of the surveillance interval. Even though the SR is not annotated with a 4.0.4 exemption, the SR is not required to be performed prior to entering a MODE in the Applicability of the associated LCO if any of the following conditions are satisfied:

(1) the SR has been performed within the surveillance interval (i.e. it is current) and is known not to be failed or (2) the SR is required to be met, but not performed, in the MODE to be entered and is known not to be failed. The initial surveillance performance will be completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> once the plant is at stable operating pressure following the establishment of steady state conditions. Other instruments such as those contained in TS 3/4.4.5.1 can be utilized to determine whether RCS operational leakage limits are being exceeded prior to initial performance.

(LBDCR 16-046, Ch. 86)

Once the plant establishes steady state operation, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for completing the SR. If the SR was not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval, there would then be a failure to perform the SR within the specified interval, and the provisions of 4.0.3 would apply. Should the interval in accordance with the Surveillance Frequency Control Program be exceeded while steady state operation has not been established, this NOTE allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after steady state operation has been established to perform the SR. The SR is still considered to be performed within the surveillance interval. Therefore, if the Surveillance was not performed in accordance with the Surveillance Frequency Control Program (plus the extension allowed by 4.0.2) interval, but steady state operation was not established, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of 4.0.4 occurs when changing MODES, even with the surveillance interval in accordance with the Surveillance Frequency Control Program not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with the establishment of steady state

< (EC-3173 Ch. 53) operation.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowable limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(LBDCR 16-046, Ch. 86)

WATERFORD - UNIT 3 B 3/4 4-4e CHANGE NO. 53, 86

REACTOR COOLANT SYSTEM.

BASES (continued)

> (DRN 04-1243, Ch. 38;06-916, Ch. 48)

The primary to secondary leakage limit of 75 gallons per day through any one SG is based on the operational leakage performance criterion in NEI 97-06. The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, AThe RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day.@ The NEI 97-06 limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion (since it is less than 150 gpd through any one SG) in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

< (DRN 04-1243, Ch. 38;06-916, Ch. 48)

OPERATIONAL LEAKAGE (Continued)

>(DRN 04-1243, Ch. 38)

Steam generator tube cracks having primary-to-secondary leakage less than 150 gpd per steam generator during operation will have an acceptable margin of safety to withstand loads imposed during normal operation and postulated accidents (Reference NEI 97-06). Due to the proximity of the east atmospheric dump valve to the east control room intake, the primary-to-secondary leakage limit required to achieve acceptable radiological consequences, for accidents that rely on reactor coolant system cooldown using the steam generators, is limiting. Therefore, 75 gpd per steam generator is imposed as the primary-to-secondary operational leakage limit.

<(DRN 04-1243, Ch. 38)

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. is prohibited as the leak itself could cause further reactor coolant pressure boundary (RCPB) deterioration, resulting in higher leakage. If PRESSURE BOUNDARY LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal leakage past the isolation device is acceptable as it will limit RCS leakage and is included in IDENTIFIED LEAKAGE or UNIDENTIFIED LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

(LBDCR 16-046, Ch. 86)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(LBDCR 16-046, Ch. 86)

>(LBDCR 13-003, Ch. 74) 3/4.4.6 DELETED

<(LBDCR 13-003, Ch. 74)

WATERFORD - UNIT 3 B 3/4 4-5 CHANGE NO. 18, 38, 39, 74, 86,

REACTOR COOLANT SYSTEM.

BASES (continued) 3/4.4.7 SPECIFIC ACTIVITY

>(DRN 03-173, Ch. 18;05-131, Ch. 39)

The Code of Federal Regulations, 10 CFR 50.67 specifies the maximum total effective dose equivalent an individual offsite can receive during a design basis accident. The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity. The specific activity limits ensure that these doses are held within the appropriate 10 CFR 50.67 requirements (small fraction, well within, or within) during analyzed transients and accidents.

<(DRN 05-131, Ch. 39)

Operation with iodine specific activity levels greater than the LCO limit is permissible for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided the activity levels do not exceed 60 uCi/gm. A 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> limit was established because of the low probability of an accident occurring during this period. The dose consequences of an accident during this 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period would not exceed the full 10 CFR 50.67 limits.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

<(DRN 03-173, Ch. 18)

(LBDCR 16-046, Ch. 86)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(LBDCR 16-046, Ch. 86)

Enclosure, Attachment 3 CNRO2021-00020 Retyped Technical Specification Pages ANO-1 ANO-2 GGNS RBS Waterford-3 Number of Plant Affected TS Pages Arkansas Nuclear One - Unit 1 3 Arkansas Nuclear One - Unit 2 3 Grand Gulf Nuclear Station, Unit 1 3 River Bend Station, Unit 1 3 Waterford Steam Electric Station, Unit 3 4 Total Number of Pages in Attachment 3 16

Enclosure, Attachment 3 CNRO2021-00020 Retyped Technical Specification Pages Arkansas Nuclear One - Unit 1 Technical Specification Affected Number TS Number Title of Pages 1.1 Definition 2 3.4.13 RCS Operational LEAKAGE 1 Total ANO-1 TS Pages 3

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion."

DOSE EQUIVALENT XE-133 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-131m, 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."

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE).

ANO-1 1.1-3 Amendment No. 215,224,243,257,270,

Definitions 1.1 1.1 Definition (continued)

LEAKAGE (continued) b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE.

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MODE 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.

OPERABLE-OPERABILITY 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, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS 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 the SAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

Power in any Core Quadrant QPT = 100 1 Average Power in all Quadrants ANO-1 1.1-4 Amendment No. 215,243,270,

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

d. 150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary A.1 Isolate affected component, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE exists. pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

B. RCS unidentified or B.1 Reduce LEAKAGE to within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> identified LEAKAGE not limits.

within limits, except for primary to secondary LEAKAGE.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND OR C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Primary to secondary LEAKAGE not within limit.

ANO-1 3.4.13-1 Amendment No. 215,224,

Enclosure, Attachment 3 CNRO2021-00020 Retyped Technical Specification Pages Arkansas Nuclear One - Unit 2 Technical Specification Affected Number TS Number Title of Pages 1.14 Definitions - IDENTIFIED LEAKAGE 1 1.16 PRESSURE BOUNDARY LEAKAGE 1 3.4.6.2 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 1 Total ANO-2 TS Pages 3

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c. Digital computer channels - The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SHUTDOWN MARGIN 1.13 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 control element assemblies are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 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 to not interfere with the operation of leakage detection systems, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

ARKANSAS - UNIT 2 1-3 Amendment No. 157,220,255,266, 319,322,

DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

AZIMUTHAL POWER TILT - Tq 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT I-131 1.18 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

DOSE EQUIVALENT XE-133 1.19 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-131m, 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.

1.20 Deleted FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

ARKANSAS - UNIT 2 1-4 Amendment No. 157,255,266,282, 293,315,

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and primary to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

d. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two valves* in each high pressure line having a non-functional valve and 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• These valves may include check valves for which the leakage rate has been verified, manual valves or automatic valves. Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.

ARKANSAS - UNIT 2 3/4 4-14 Amendment No. 184,266,280, Order dated 4/20/81

Enclosure, Attachment 3 CNRO2021-00020 Retyped Technical Specification Pages Grand Gulf Nuclear Station, Unit 1 Technical Specification Affected Number TS Number Title of Pages 1.1 Definitions - LEAKAGE 1 3.4.5 RCS Operational LEAKAGE 2 Total GGNS TS Pages 3

Definitions 1.1 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length RATE (LHGR) of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may (continued)

GRAND GULF 1.0-4 Amendment No. 120,

RCS Operational LEAKAGE 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Operational LEAKAGE LCO 3.4.5 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 5 gpm unidentified LEAKAGE;

c. 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d. 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary A.1 Isolate affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE exists. component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

B. Unidentified LEAKAGE not B.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within limit. within limits.

OR Total LEAKAGE not within limit.

(continued)

GRAND GULF 3.4-12 Amendment No. 120,

RCS Operational LEAKAGE 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Unidentified LEAKAGE C.1 Verify source of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> increase not within limit. unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY In accordance with SR 3.4.5.1 Verify RCS unidentified LEAKAGE, total LEAKAGE, the Surveillance and unidentified LEAKAGE increase are within Frequency Control limits.

Program GRAND GULF 3.4-13 Amendment No. 120, 219,

Enclosure, Attachment 3 CNRO2021-00020 Retyped Technical Specification Pages River Bend Station, Unit 1 Technical Specification Affected Number TS Number Title of Pages 1.1 Definitions - LEAKAGE 1 3.4.5 RCS Operational LEAKAGE 2 Total RBS TS Pages 3

Definitions 1.1 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

(continued)

RIVER BEND 1.0-4 Amendment No. 81,

RCS Operational LEAKAGE 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Operational LEAKAGE LCO 3.4.5 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 5 gpm unidentified LEAKAGE;

c. 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d. 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary A.1 Isolate affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE exists. component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

B. Unidentified LEAKAGE not B.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within limit. within limits.

OR Total LEAKAGE not within limit.

(continued)

RIVER BEND 3.4-12 Amendment No. 81, 196,

RCS Operational LEAKAGE 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Unidentified LEAKAGE C.1 Verify source of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> increase not within limit. unidentified LEAKAGE increase is not service sensitive type 304, type 316 austenitic stainless steel, or other intergranular stress corrosion cracking susceptible material.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify RCS unidentified LEAKAGE, total LEAKAGE, In accordance with and unidentified LEAKAGE increase are within limits. the Surveillance Frequency Control Program RIVER BEND 3.4-13 Amendment No. 81, 196,

Enclosure, Attachment 3 CNRO2021-00020 Retyped Technical Specification Pages Waterford Steam Electric Station, Unit 3 Technical Specification Affected Number TS Number Title of Pages 1.14 DEFINITIONS - IDENTIFIED LEAKAGE 1 1.21 & DEFINITIONS - PRESSURE BOUNDARY LEAKAGE 1 1.23 DEFINITIONS - PURGE - PURGING 3.4.5.2 OPERATIONAL LEAKAGE 2 Total Waterford-3 TS Pages 4

DEFINITIONS IDENTIFIED LEAKAGE (Continued)

b. Leakage into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems, or

c. Reactor Coolant System leakage through a steam generator to the secondary system (primary to secondary leakage).

MEMBER(S) OF THE PUBLIC 1.15 MEMBER(S) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.16 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specification 6.9.1.7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.17 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).

OPERATIONAL MODE - MODE 1.18 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.

WATERFORD - UNIT 3 1-4 AMENDMENT NO. 68, 84, 116, 204,

DEFINITIONS PHYSICS TEST 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (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.

PLANAR RADIAL PEAKING FACTOR - Fxy 1.20 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall, or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that process-ing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be 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.

WATERFORD - UNIT 3 1-5 AMENDMENT NO. 68, 204,

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 75 gallons per day primary to secondary leakage, through any one steam generator (SG),

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 1 gpm leakage at a Reactor Coolant System pressure of 2250 +/- 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With any primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System operational leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE, primary to secondary leakage, and leakage from Reactor Coolant System pressure isolation valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

d. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

WATERFORD - UNIT 3 3/4 4-18 AMENDMENT NO. 197,199, 204, 249,

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS NOTE: Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

4.4.5.2.1 Reactor Coolant System leakages, except for primary to secondary leakage, shall be demonstrated to be within each of the above limits by performance of a Reactor Coolant System water inventory balance in accordance with the Surveillance Frequency Control Program.

4.4.5.2.2 Primary to secondary leakage shall be verified to be 75 gallons per day through any one SG in accordance with the Surveillance Frequency Control Program.

4.4.5.2.3 Each Reactor Coolant System pressure isolation valve specified in Table 3.4-1, Section A and Section B, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. In accordance with the Surveillance Frequency Control Program,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve,

d. Following valve actuation for valves in Section B due to automatic or manual action or flow through the valve:

1. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying valve closure, and

2. Within 31 days by verifying leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.5.2.4 Each Reactor Coolant System pressure isolation valve power-operated valve specified in Table 3.4-1, Section C, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. In accordance with the Surveillance Frequency Control Program, and

b. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

WATERFORD - UNIT 3 3/4 4-19 AMENDMENT NO. 96,107, 204, 249,