ML061560144

Response to NRC Request for Additional Information Re License Amendment Request 218: Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

ML061560144
Person / Time
Site:	Kewaunee
Issue date:	06/02/2006
From:	Grecheck E Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
05-562A
Download: ML061560144 (86)

Text

Dominion Energy Kewaunee, Inc.

i 0 0 0 Dominion Roulcvard, Glen Allen, VA 2.1060 p Dominion" June 2 , 2006 U. S. Nuclear Regulatory Commission Serial No. 05-562A Attention: Document Control Desk KPS/LIC/GR: R1 Washington, D.C. 20555 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE. INC.

KEWAUNEE POWER STATION RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 218: APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY In accordance with 10 CFR 50.90, Dominion Energy Kewaunee, Inc. (DEK) submitted a request for an amendment to the technical specifications (TS) for Kewaunee Power Station (Kewaunee) (Reference 1). Subsequent to DEK's submittal, the NRC staff requested additional information to complete its review. provides the questions asked by the NRC staff with DEK's responses. provides a complete replacement of the existing TS pages marked-up to show the proposed changes. Attachment 3 provides a complete replacement of the proposed TS pages. Attachments 4 and 5 provide a complete replacement of the marked-up and proposed TS Basis pages for information only. Complete replacement of the TS and TS Basis pages are being provided to avoid confusion concerning which pages have been changed because of the request for additional information (RAI). The RAI responses do not change the significant hazards determination for the proposed amendment discussed in Reference 1.

As stated in Reference 1, DEK requests approval of the proposed amendment by June 30, 20061 to facilitate scheduling refueling activities. Once approved, DEK will implemenit this amendment within 60 days.

In accordance with 10 CFR 50.91 (b), a copy of this letter, with attachments, is being provided to the designated Wisconsin official.

If you have any questions, please contact Mr. Gerald Riste at 920-388-8424.

Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services

Serial No. 05-562A KPS LAR 218 SG CLllP Page 2 of 4 Commitmlents made in this letter: None Reference

1. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), "License Amendment Request 218 - Application For Technical Specification lmprovement Regarding Steam Generator Tube Integrity," dated January 12, 2006. (ADAMS Accession No. ML060250524)

Attachments

1. Response to NRC Request for Additional Information Regarding License Amendment Request - 218, Application for Technical Specification lmprovement Regarding Steam Generator Tube Integrity

2. Marked Up Proposed Technical Specification Pages

3. Proposed Technical Specification Pages

4. Marked Up Technical Specification Bases Pages

5. Proposed Technical Specification Bases Pages

Serial No. 05-562A KPS LAR 218 SG CLllP Page 3 of 4 cc: Regional Administrator U. S. Nuclear Regulatory Commission Region Ill 2443 Warrenville Road Suite 210 Lisle, Illinois 60532-4352 Mr.. D. H. Jaffe Project Manager U.S. Nuclear Regulatory Commission Mail Stop 0-7-D-1 Washington, D. C. 20555 Mr. S. C. Burton NFlC Senior Resident Inspector Kewaunee Power Station Public Service Commission of Wisconsin Electric Division P.O. Box 7854 Madison, WI 53707

Serial No. 05-562A KPS LAR 218 SG CLllP Page 4 of 4 COMMONWEALTH OF VIRGINIA )

1 COUNTY OF HENRICO 1 The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is the Vice President -

Nuclear Support Services of Dominion Energy Kewaunee, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 2 37 day of ,2006.

My Commission Expires: (

Notary Public

, 2? J & M ,

(SEAL)

ATTACHMENT 1 DOMINION ENERGY KEWAUNEE, INC. RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST -

2118, APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 1 of 10 Dominion Enerav Kewaunee's Res~onseto NRC's Reauest for Additional Informatiion Reaardina License Amendment Reauest - 218, A~plication for Technical S~ecificationlm~rovementReaardina Steam Generator Tube lntearity The Nuclear Regulatory Commission (NRC) has requested additional information regarding Operating License DPR-43, License Amendment Request (LAR) 21 8, "Application For Technical Specification Improvement Regarding Steam Generator Tube Integrity," for Kewaunee Power Station (Kewaunee).

Each of the NRC staff's questions (Reference 3) and DEK's corresponding response are provided as follows:

NRC Question 1:

According to your proposed TS 3.1 .d, the operational primary-to-secondary leakage limit will be 150 gallons per day (gpd) per steam generator (SG), which is the value approved in TSTF-449. According to Page TS 83.1 -17 of your proposed Bases, the accident induced leakage limit for design basis accidents (DBAs) other than a steam generator tube rupture will also be 150 gpd per SG. During a main steam line break (MSLB) accident, which is one of your postulated DBAs, the differential pressure across the tubes is greater than the differential pressure during normal operation.

Therefore, the primary-to-secondary leakage may be greater during a MSLB than during normal operation. Since you could be operating with leakage as high as your normal operating leakage limit (150 gpd per SG), the amount of leakage during some postulated accidents could be greater than that assumed in your accident analyses.

Please disscuss what controls are in place to ensure that you do not exceed your accident induced leakage performance criteria as a result of operational leakage (i.e.,

operational leakage at or below the TS limit).

DEK Response:

Dominion Energy Kewaunee, Inc. (DEK) discussed these controls in license amendment request (LAR) 21 8. Starting on page 2 of 9 of Attachment 1 DEK stated:

"At Kewaunee, installed Radiation Monitoring Systems (RMSs) pro vide continuous online monitoring of primary-to-secondary leakage to plant operators. Kewaunee operating procedure E 14, Steam Generator Tube Leak, provides actions to take when a small primary-to-secondary steam generator tube leak exists. A small tube leak is defined as one that is great~erthan 5 gallons per day in any steam generator. The procedure requires confirmation and monitoring of the leak rate to determine if the leak has stabilized. Operations, Engineering, and Radiation Protection are

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 2 of 10 notified of the condition and participate in the evaluation and monitoring of the situation.

If the leak rate increases to 30 gallons per day, E-0-14 directs the operators to place the secondary radiation monitors on continuous trend, monitor every 15 minutes, and verify the secondary radiation monitors alarm setpoints. E-0- 14 directs chemistry to increase the grab sampling frequency, determine which steam generator is leaking, and determine the new leakrate.

If primary-to-secondary leakage is 75 gpd or greater for greater than one hour, the operators place the secondary radiation monitors on continuous trend and monitor every 15 minutes. Actions are initiated to perform a normal plant shutdown and achieve the Hot Shutdown condition (reactor shutdown and RCS Tavg greater than or equal to 540 OF) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If primary-to-secondary leakage is 100 gpd or greater, a rapid plant shutdown is initiated. E 14 directs the operators to reduce plant power to less than 50% within one hour and requires that the plant be placed in the Hot Shutdown condition within the next two hours."

These controls are in place to ensure that Kewaunee does not exceed its accident induced leakage performance criteria due to operational primary-to-secondary leakage.

NRC Question 2:

Page TS B3.1-7 of your proposed Bases states the following regarding accident induced leakage:

"The accident analysis assumes that accident induced leakage does not exceed 150 gallons per day per SG, except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage."

Since you do not have any degradation for which the NRC has approved greater accident induced leakage, please discuss your plans to modify your Bases to remove the last part of this sentence.

DEK Res~~onse:

The proposed TS Basis pages have been modified to delete the phrase, "except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage." Kewaunee does not have any degradation for which the

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 3 of 10 NRC has approved greater accident induced leakage. The modified TS Basis pages are included in Attachment 4, "Marked Up Technical Specification Bases Pages," and , "Proposed Technical Specification Bases Pages."

NRC Question 3:

The proposed Bases sections TS 3.1 .d.2 and TS 3.1 .d.3 are inconsistent with TS 3.1.d and TSTF-449 with respect to the actions required if primary to secondary leakage exceeds the limiting condition for operation. TSTF-449 and your proposed TS 3.1.d require th~eplant to be brought to a lower-pressure operating mode (HOT SHUTDOWN for Kewaunee) if primary to secondary leakage exceeds the limit. Your proposed Bases sections TS 3.1 .d.2 and TS 3.1 .d.3 (page TS B3.1-12) state that if primary to secondary leakage is above the TS limit, you would be allowed four hours to reduce this leaka.geto within the limit without having to initiate action to achieve HOT SHUTDOWN. Please discuss your plans for modifying these proposed TS Bases to be consistent with your proposed TS and TSTF-449.

DEK Res~~onse:

DEK agre!es with the NRC staff and has modified the proposed TS Basis. The modified TS Basis pages are included in Attachment 4, "Marked Up Technical Specification Bases Pages," and Attachment 5, "Proposed Technical Specification Bases Pages."

NRC Que!stion 4:

Proposed surveillance requirements (SR) for RCS Operational Leakage (TS 4.1 8) and Steam Generator (SG) Tube lntegrity (TS 4.1 9) state that these SR are applicable to surveillance requirements for RCS Operational Leakage and Steam Generator Tube Integrity. However, these TS do not refer to the corresponding limiting conditions for operation (LCO), Sections 3.1 .d (RCS Operational Leakage) and 3.1 .g (Steam Generator (SG) Tube Integrity). Other sections in your existing TS (e.g., TS 4.1 7, Control Room Post-Accident Recirculation System) do refer to the corresponding LC0 section. Please discuss your plans for modifying your proposed TS in order to be consistent with your existing TS and to relate the applicability of these SR to the correspor~dingLCO.

DEK Response:

DEK has modified the surveillance specification for RCS Operational Leakage (TS 4.18) and Steam Generator Tube (SG) Integrity (TS 4.19) to reference TS sections 3.1 .d and 3.1 .g, respectively. These modifications are included in Attachment 2,

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page4of 10 "Marked Up Technical Specification Pages," and Attachment 3, "Proposed Technical Specification Pages."

NRC Question 5:

In your marked-up Bases, in the section on Leakage of Reactor Coolant (page B3.1-1O), you indicate the steam line break is less limiting for site radiation releases.

However, a preceding paragraph indicates that other accidents (such as a steam generator tube rupture) involves secondary steam releases to a lesser extent than the SLB accident. Please clarify which accident is the most limiting for site radiation releases 'and discuss your plans to incorporate this into your Bases.

DEK Reslponse:

The Radiological Accident Analysis (RAA) of record for Kewaunee identifies the Locked Rotor accident as the most limiting for site radiation releases. The Steam Generator Tube Rupture (SGTR) is next, followed by the Control Rod Ejection Accident, and the Steam Line Break (Reference 1). DEK has recently submitted to the NRC a revision to the Kewaunee RAA which identifies the SGTR as the most limiting for site radiation releases, followed by the locked rotor and rod ejection accidents, and finally the steam line break (Reference 2). Information associated with these accidents have been added to the TS basis and is included in Attachment 4, "Marked Up Technical Specification Bases Pages," and Attachment 5, "Proposed Technical Specification Bases Pa.ges."

NRC Question 6:

Proposed TS 3.1 .d.l .D uses the abbreviation "SG" without defining it. The corresporiding item (LC0 3.4.1 3.d) in TSTF-449 is:

"150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG)"

Please di:scussyour plans for modifying the proposed TS to make it consistent with TSTF-44!3.

DEK Response:

DEK has modified TS 3.1 .d.l .D to define "SG." The modified TS pages are included in , "Marked Up Technical Specification Pages," and Attachment 3, "Proposed Technical Specification Pages."

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 5 of 10 NRC Question 7:

The format and wording of proposed TS 3.1 .g.2 and TS 3.1 .g.3 differ from the TSTF.

Based on the staff's understanding of the intent of the TSTF, the words "were not" (for TS 3.1 .g.2) and "are not" (for TS 3.1 .g.3) would be more appropriate than the term "can not be" in your proposed TS (e.g., "If the requirements of TS 3.1 .g.l .B were not met, then:"). Please discuss your plans for modifying the proposed TS to address this comment.

DEK Response:

DEK has reviewed TS 3.1 .g.2, TS 3.1 .g.3, and TSTF 449, Rev 4, and agrees with the NRC staff that the wording of these TS items can be improved to better meet the intent of TSTF 449, Rev 4. DEK proposes that in both TSs the phrase "can not be" would be more appropriately changed to "are not." This shows that both actions reflect the present condition of the steam generator tube(s). Therefore, these specifications would not apply to SG tubes that, in the past, were not plugged but may now be plugged.

Rather, the specification would apply to those tubes that should be plugged and currently are not.

DEK has modified TS 3.1 .g.2 and TS 3.1 .g.3 to change "can not be" to "are not." The modified 'TS pages are included in Attachment 2, "Marked Up Technical Specification Pages," and Attachment 3, "Proposed Technical Specification Pages."

NRC Question 8:

In proposed TS 3.1 .g.3, the word "to" is missing after "initiate action" (i.e., "If the requirements of TS 3.1 .g.2.A or TS 3.1 .g.l .A are not met, then initiate action to:").

Adding the word "to" would make this TS consistent with similar statements in proposed TS 3.1 .d. Please discuss your plans for modifying the proposed TS to address this comment.

DEK Response:

DEK agrees with the proposed change. TS 3.1 .g.3 has been modified to add the word "to" after the phrase, "initiate action." The modified TS pages are included in , "Marked Up Technical Specification Pages," and Attachment 3, "Proposed Technical Specification Pages."

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page6of 10 NRC Question 9:

The term "REFUELING SHUTDOWN" is used in the proposed TS Bases on pages TS B3.1-10 and TS B3.1-19. This mode is not defined in the table of Kewaunee operational modes on page 1 of Attachment 2 in the license amendment request.

Please discuss your plans to modify your proposed TS Bases to address this inconsistency.

DEK Reslponse:

The word "SHUTDOWN" has been deleted and the mode identified as "REFUELING".

The TS Basis pages have been modified and are included in Attachment 4, "Marked Up Technicall Specification Basis Pages," and Attachment 5, "Proposed Technical Specification Basis Pages."

NRC Question 10:

The staff observes that throughout the proposed TS Bases, the limiting conditions for operation (LCO) are referred to as "TS requirements." For example, on page TS B3.1-10, under TS 3.1 .d.l , the paragraph under "A. Pressure Boundary LEAKAGE," reads, "Violation of this TS requirement could result." However, in the first sentence of the Basis secition on Operational Leakage (TS 4.1 8), you refer to "the TS LC0 limits."

Please discuss your plans to modify your proposed TS Bases to address this inconsistency.

DEK Response:

Kewaunele has custom technical specifications (CTS) that pre-date NUREG 0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors,"

(STS) and NUREG 1431, "Standard Technical Specifications Westinghouse Plants,"

(ISTS) which leads to some confusion.

10CFR50.36(~)(2)defines Limiting Conditions for Operations as:

"Limiting conditions for operation are the lowest functional capability or perhrmance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met."

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 7 of 10 This definition and usage is clearly identified in ISTS where limiting conditions for operation (LCO) has this designation given. In ISTS, each system, structure, or component (SSC) that 10CFR50.36 requires to be in TS has an LC0 specifically labeled as the LCO. Kewaunee CTS do not have this feature.

In Kewaunee CTS, TS l.O.d defines Limiting Conditions For Operation as:

"Limiting conditions for operation are those restrictions on reactor operation, resulting from equipment performance capability that must be enforced to ensure safe operation of the facility."

Comparing the 950.36 LC0 description with Kewaunee CTS Section 1.O definition of LC0 reveals that the Kewaunee CTS definition includes the 950.36 description of an LC0 and the actions to be performed when an LC0 is not met. Additionally, in the Kewaunele CTS each listed SSC is divided into three sections: Applicability, Objective, and Specifications.

Therefore!, because Kewaunee CTS do not specifically identify any item in the Limiting Conditions for Operation section (Section 3.0) as an LC0 (except for TS 3.1 .a.3), and the items referred to are in the Specifications section, DEK will modify the Kewaunee TS Bases included in Reference 1 to change "LCO" and "requirements" to "specification(s)."

NRC Question 11:

TS 3.1 .d.2 should probably specifically exclude primary-to-secondary leakage and pressure lboundary leakage. This would be consistent with the TSTF. The reason is that primary-to-secondary leakage is considered identified leakage and it could lead to confusion with 3.1 .d.3.

DEK Response DEK agrees with the proposed change. TS 3.1.d.2 has been modified to specifically exclude primary-to-secondary leakage and pressure boundary leakage. The modified TS pages are included in Attachment 2, "Marked Up Technical Specification Pages,"

and Attachment 3, "Proposed Technical Specification Pages."

NRC Question 12:

The licensee may want to add "(SG)" after "Steam Generator" in the title for TS 3.1.g.

This will make it consistent with the TSTF.

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 8 of 10 DEK Reslponse:

DEK agrebes with the proposed change. TS 3.1 .g has been modified to add "(SG)" after "Steam Generator" in the title. The modified TS pages are included in Attachment 2, "Marked Up Technical Specification Pages," and Attachment 3, "Proposed Technical Specification Pages." Similarly, DEK proposes to add "(SG)" after "Steam Generator" in the title! for TS 4.1 9, this TS contains the surveillance requirements supporting TS 3.1 .g.

NRC Question 13:

TS Bases page B3.1-18: In the 2nd sentence of the 2nd paragraph under Applicable Safety Analysis, the licensee should probably add "...or is assumed to increase to 300 gpd as a result of accident induced leakage." This will make their submittal consistent with the T'STF (and with the definition of accident induced leakage).

DEK Response:

DEK agrees with the proposed change. TS Bases page B3.1-18 has been modified to

"...or is assumed to increase to 300 gpd as a result of accident induced leakage." The modified TS Basis pages are included in Attachment 4, "Marked Up Technical Specification Basis Pages," and Attachment 5, "Proposed Technical Specification Basis Pages."

Additional Chanaes TS Basis Com~letionTime Descri~tion Based on DEK's re-review of the LAR submittal, an additional change is being proposed. In the TS Basis for TS 3.1.d, "RCS Operational LEAKAGE," and TS 3.1 .g, "Steam Generator (SG) Tube Integrity," the completion time description for entering Cold Shutdown did not match the corresponding Technical Specification.

In TSTF 4149, Revision 4, Specification 3.4.13, "RCS Operational Leakage," and Specification 3.4.20, "Steam Generator (SG) Tube Integrity," Action "B" the total completion time for placing the reactor in Cold Shutdown is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Kewaunee proposed TS 3.1 .d.3 states:

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page9of 10 "If the limits contained in TS 3.1.d. I for pressure boundary or primary to secondary LEAKAGE are exceeded, or the time limit contained in TS 3.7 .d.2 is exceeded, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. "

Kewaunee proposed TS 3.1 .g.3 states:

"If the requirements of TS 3.1.g.2.A or TS 3.1 .g. 1.A are not met, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."

Kewaunee proposed TS Basis for these action items completion times stated:

"The reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after achieving HOT SHU'TDOWN. "

This statement is in conflict with current Kewaunee TS and TSTF 449. As with the TSTF 449, Revision 4 required action and completion time, once the operators enter TS 3.1 .d.3 or TS 3.1 .g.3, 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is allowed to place the reactor in the Cold Shutdown condition.

Therefore, the proposed Basis for TS 3.1 .d.3 and TS 3.1 .g.3 has been modified to state :

"The reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."

Title Corrlections

1. In the Table of Contents, DEK proposes to change certain TS titles to match the titles of the specifications. These title changes are as follows:

3.1 .d, changed from "Leakage of Reactor Coolant" to "RCS Operational Leakage."

3.1 .g, added "(SG)" after Steam Generator.

Serial No. 05-562A Docket No. 50-305 Attachment 1 Page 10 of 10 4.1 9, added "(SG)" after Steam Generator.

6.22, added "(SG)" after Steam Generator.

2. In the TS Bases title for TS 3.1 .dl DEK proposes to change the title from "Leakage of Reactor Coolant (TS 3.1 .d)" to "RCS Operational LEAKAGE (TS 3.1 .d)."

3. In the TS Bases title for TS 4.1 8, proposes to capitalize the word "Leakage" as it is a TS defined term.

4. In the TS Bases title for TS 4.1 9, DEK proposes to add "(SG)" after Steam Generator.

1. Letter from John G. Lamb (NRC) to Thomas Coutu (NMC), "Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Implementation of Alternate Source Term (TAC NO. MB4596)," dated March 17,2003. (ADAMS Accession NO.

MLO3O210062).

2. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), "License Amendment Request 21 1 - Radiological Accident Analysis and Associated Technical Specifications Change," dated January 30, 2006. (ADAMS Accession No. MI-06054021 7).

3. E-mails from D. H. Jaffe (NRC) to G.O. Riste (DEK) dated April 21, 2006, and May 15, 2006.

ATTACHMENT 2 LICENSE AMENDMENT REQUEST 218 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY MARKED UP PROPOSED TECHNICAL SPECIFICATION PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

TABLE OF CONTENTS TECHNICAL SPECIFICATIONS APPENDIX A Section Title Page 1.0 Definitions ............................................................................................................... 1.0-1 Quadrant-to-Average Power Tilt Ratio .........................................................1.0.1 Safety limits................................................................................................ 1.0.1 Limiting Safety System Settings ..................................................................1.0.1 Limiting Conditions for Operation.................................................................1.0.1 Operable .Operability ................................................................................1.0.1 Operating .................................................................................................. 1.0.1 Containment System Integrity...................................................................... 1.0.2 Protective Instrumentation Logic .................................................................1.0.2 Instrumentation Surveillance ....................................................................... 1.0.3 Modes ....................................................................................................... 1.0.4 Reactor Critical........................................................................................... 1.0.4 Refueling Operation ...................................................................................1.0.4 Rated Power .............................................................................................. 1.0.4 Reportable Event .......................................................................................1.0.4 Radiological Effluents ................................................................................. 1.0.5 Dose Equivalent 1-131 ................................................................................ 1.0.6 Core Operating Limits Report ...................................................................... 1.0.6 Shutdown Margin . ....................................................................................... 1.0.6 Immediately................................................................................................ 1.0.6 1.0.t Leakaae .................................................................................................... 1-0-7 1 2.0 Safety Limits and Limiting Safety System Settings .....................................................2.1-1 2.1 Safety Limits. Reactor Core ........................................................................ .2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure........................................... 2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation ..........................................................................................2.3-1 2.3.a Reactor Trip Settings ................................................................2.3-1 2.3.a.l Nuclear Flux ........................................................2.3-1 2.3.a.2 Pressurizer..........................................................2.3-1 2.3.a.3 Reactor Coolant Temperature ............................. 2.3-2 2.3.a.4 Reactor Coolant Flow .......................................... 2.3-3 2.3.a.5 Steam Generators .............................................. .2.3-3 2.3.a.6 Reactor Trip Interlocks ........................................2.3-4 2.3.a.7 Other Trips ..........................................................2.3-4 3.0 Limiting Conditions for Operation.............................................................................. 3.0.1 3.1 Reactor Coolant System ............................................................................ .3. 1-1 3.1 .a Operational Components...........................................................3. 1-1 3.1 .a.1 Reactor Coolant Pumps ......................................3.1-1 3.1 .a.2 Decay Heat Removal Capability...........................3.1-1 3.1 .a.3 Pressurizer Safety Valves ................................... .3. 1-3 3.1 .a.4 Pressure Isolation Valves ....................................3.1-4 3.1 .a.5 Pressurizer PORV and PORV Block Valves .........3.1-4 3.1 .a.6 Pressurizer Heaters.............................................3.1-5 3.1 .a.7 Reactor Coolant Vent System ............................. .3.1-5 3.1 .b Heatup & Cooldown Limit Curves for Normal Operation............. 3.1-6 3.1.c Maximum Coolant Activity ......................................................... 3.1-7 3.1 .d R-CS Operational Leakaae..............3.1-8 1 3.1 .e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration .............................................................. 3.1-9 3.1.f Minimum Conditions for Criticality............................................ 3.1-10 3.1.a Steam Generator (SG) Tube Intearitv ......................................3.1-11 1 LAR 218 TS i

Section Title 7

Paue 3.2 Chemical and Volume Control System............................................................ 3.2.1 3.3 Engineered Safety Features and Auxiliary Systems ........................................ 3.3.1 3.3.a Accumulators ...........................................................................3.3.1 3.3.b Emergency Core Cooling System ..............................................3.3-2 3.3.c Containment Cooling Systems................................................... 3.3-4 3.3.d Component Cooling System ......................................................3.3-6 3.3.e Service Water System ............................................................... 3.3-7 Steam and Power Conversion System......................................................... 3.4-1 3.4.a Main Steam Safety Valves ........................................................ 3.4-1 3.4.b Auxiliary Feedwater System ...................................................... 3.4-1 3.4.c Condensate Storage Tank ........................................................ .3.4-3 3.4.d Secondary Activity Limits ...........................................................3.4-3 Instrumentation System .............................................................................. 3.5-1 Containment System .................................................................................. 3.6-1 Auxiliary Electrical Systems ......................................................................... 3.7-1 Refueling Operations .................................................................................

.3.8-1 Deleted Control Rod and Power Distribution Limits ................................................. 3.10-1 3.1O.a Shutdown Reactivity................................................................ 3.10-1 3.10.b Power Distribution Limits ......................................................... 3.10-1 3.10.c Quadrant Power Tilt Limits ...................................................... 3.10-4 3.1O.d Rod Insertion Limits ................................................................ 3.10-4 3.1O.e Rod Misalignment Limitations .................................................. 3.10-5 3.1O.f Inoperable Rod Position Indicator Channels ............................3.10-5 3.10.9 Inoperable Rod Limitations ......................................................3.10-7 3.1O.h Rod Drop Time ........................................................................ 3.10-7 3.10.i Rod Position Deviation Monitor................................................3.10-7 3.10.j Quadrant Power Tilt Monitor................................................... .3.10-7 3.1O.k Core Average Temperature .....................................................

3.10-7 3.10.1 Reactor Coolant System Pressure ...........................................3.10-7 3.1O.m Reactor Coolant Flow .............................................................3.10-8 3.1O.n DNBR Parameters................................................................... 3.10-8 Core Surveillance Instrumentation ............................................................. 3.1 1-1 Control Room Post-Accident Recirculation System .................................... 3.12-1 Shock Suppressors (Snubbers) ................................................................. 3.14-1 4.0 Surveillance Requirements.......................................................................................4.0.1 4.1 Operational Safety Review ........................................................................ 4.1.1 4.2 ASME Code Class In-service Inspection and Testing .................................. 4.2.1 4.2.a ASME Code Class 1 , 2 , 3, and MC Components and Supports ..................................................................................4.2-1 4.2.b Deleted ........ .4.2-2Ska.r~Gcs- ..............................? .2 2 4.2.b.6 4.2.b.7 8 . U A 9-45

..L 4.3 Deleted

Section Title 4.4 Containment Tests 4.4.1 4.4.a Integrated Leak Rate Tests (Type A) ......................................... 4.4-1 4.4.b Local Leak Rate Tests (Type B and C) ......................................4.4-1 4.4.c Shield Building Ventilation System.............................................4.4-1 4.4.d Auxiliary Building Special Ventilation System ............................. 4.4-3 4.4.e Containment Vacuum Breaker System ......................................4.4-3 4.4.f Containment Isolation Device Position Verification .............................................................................. .4.4-3 4.5 Emergency Core Cooling System and Containment Air Cooling System Tests .............................................................................. .4.5.1 4.5.a System Tests ...........................................................................4.5-1 4.5.a.l Safety Injection System .......................................4.5-1 4.5.a.2 Containment Vessel Internal Spray System ................................................................ 4.5-1 4.5.a.3 Containment Fan Coil Units ................................. 4.5-2 4.5.b Component Tests ......................................................................4.5.2 4.5.b.l Pumps ................................................................4.5.2 4.5.b.2 Valves ................................................................ .4.5-2 4.6 Periodic Testing of Emergency Power System............................................. 4.6-1 4.6.a Diesel Generators .................................................................... .4.6-1 4.6.b Station Batteries........................................................................ 4.6-2 4.7 Main Steam Isolation Valves ....................................................................... 4.7-1 4.8 Auxiliary Feedwater System ........................................................................ 4.8-1 4.9 Reactivity Anomalies .................................................................................. 4.9-1 4.110 Deleted 4.1 1 Deleted 4.1.2 Spent Fuel Pool Sweep System ................................................................ 4.12-1 4.1 3 Radioactive Materials Sources .................................................................. 4.1 3-1 4.1.4 Testing and Surveillance of Shock Suppressors (Snubbers)......................4.1 4-1 4.1.5 Deleted 4.115 Reactor Coolant Vent System Tests .......................................................... 4.16-1 4.1'7 Control Room Postaccident Recirculation System ..................................... 4.17-1 4.18 RCS Operational Leakaae ........................................................................ .4.18-1 4.1!9 Steam Generator (SG) Tube lntegritv ........................................................ 4.1 9-1 5.0 Design Features ......................................................................................................5.1-1 5.1 Site ........................................................................................................... 5.1-1 5.2 Containment ...............................................................................................5.2-1 5.2.a Containment System ................................................................. 5.2-1 5.2.b Reactor Containment Vessel ....................................................5.2.2 5.2.c Shield Building .......................................................................... 5.2-2 5.2.d Shield Building Ventilation System.............................................5.2-2 5.2.e Auxiliary Building Special Ventilation Zone and Special Ventilation System ........................................................5.2-2 5.3 Reactor Core ............................................................................................. .5.3.1 5.3.a Fuel Assemblies ....................................................................... .5.3-1 5.3.b Control Rod Assemblies ............................................................5.3-1 5.4 Fuel Storage .............................................................................................. 5.4-1 5.4.a Criticality ................................................................................. .5.4.1 5.4.b Capacity .................................................................................. .5.4-1 5.4.c Canal Rack Storage ..................................................................5.4-1 LAR 218 TS iii

Section Title Paqe 6.0 Adrninistrative Controls ............................................................................................ 6.1.1 Responsibility

............................................................................................. 6.1 -1 Organ~zat~on .............................................................................................. .6.2-1 6.2.a Off-Site Staff ............................................................................ 6.2-1 6.2.b Facility Staff ............................................................................. 6.2-1 6.2.c Organizational Changes ........................................................... .6.2-1 Plant Staff Qualifications ............................................................................ 6.3-1 Training ..................................................................................................... 6.4-1 Deleted ........................................................................................ 6 . 5 1 - 6.5-6 Deleted .....................................................................................................6.6-1 Safety Limit Violation .................................................................................. 6.7-1 Procedures ................................................................................................6.8-1 Reporting Requirements............................................................................. 6.9-1 6.9.a Routine Reports ........................................................................ 6.9.1 6.9.a.l Startup Report .................................................... .6.9-1 6.9.a.2 Annual Reporting Requirements .......................... 6.9-1 6.9.a.3 Monthly Operating Report ....................................6.9-3 6.9.a.4 Core Operating Limits Report ............................. 6.9-3 6.9.b Unique Reporting Requirements................................................ 6.9-6 6.9.b. 1 Annual Radiological Environmental Monitoring Report ................................................ 6.9-6 6.9.b.2 Radioactive Effluent Release Report ...................6.9-6 6.9.b.3 Special Reports .................................................. .6.9-6 6.9.b.4 Steam Generator Tube Inspection Report ...........6.9-6 1 Record Retention ..................................................................................... 6.10-1 Radiation Protection Program................................................................... .6.11-1 System Integrity........................................................................................ 6.12-1 High Radiation Area ................................................................................. 6.13-1 Deleted .................................................................................................. .6.14-1 Secondary Water Chemistry ...................................................................... 6.1 5-1 Radiological Effluents ...............................................................................6.16.1 Process Control Program (PCP) ................................................................ 6.17-1 Offsite Dose Calculation Manual (ODCM) .................................................. 6.18-1 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems ......................................................... 6.19-1 Containment Leakage Rate Testing Program ............................................ 6.20.1 Technical Specifications (TS) Bases Control Program ...............................6.21 -1 6.22 Steam Generator (SG) Proaram ................................................................ 6.22.1 1 718.0 Deleted LAR 218

LIST OF TABLES TABLE TITLE 1.O-1 ................. Frequency Notations 3.1-1 ................. Deleted 3.1-2 ................. Reactor Coolant System Pressure Isolation Valves 3.5-1 ................. Engineered Safety Features Initiation Instrument Setting Limits 3.5-2 ................. Instrument Operation Conditions for Reactor Trip 3.5-3 ................. Emergency Cooling 3.5-4 ................. Instrument Operating Conditions for Isolation Functions 3.5-5 ................. Instrument Operation Conditions for Safeguards Bus Power Supply Functions 3.5-6 .................Accident Monitoring Instrumentation Operating Conditions for Indication 4.1-1 ................. Minimum Frequencies for Checks, Calibrations and Test of lnstrument Channels 4.1 -2 ................. Minimum Frequencies for Sampling Tests 4.1-3 ................. Minimum Frequencies for Equipment Tests 4.2-1 ................. Deleted 4.2-2 ................. D e l e t e d 4 4.2-3 ................. Deleted LAR 218

LE!AKAGE

- shall be:

a. Identified LEAKAGE

1. LEAKAGE. such as that from pump seals or valve packina ( e x c e ~reactor t

coolant pump (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 specificallv located and known either not to interfere with the operation of leakage detection svstems or not to be pressure boundary LEAKAGE, or

3. Reactor Coolant Svstem (RCS) LEAKAGE through a steam aenerator to the Secondary Svstem (primary to secondary LEAKAGEI;

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

c. Pressure Boundary Leakage LEAKAGE (except primarv to secondary LEAKAGE) throuah a nonisolable fault in an RCS component bodv, pipe wall. or vessel wall.

LAR 218

8- C'E Sl

&.When the reactor is critical and above 2% power, two reactor coolant leak detection

- I systems of different operating principles shall be in operation with one of the two systems sensitive to radioactivity. Either system may be out of operation for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provided at least one system is OPERABLE.

LAR 218

a s t e a m Generator (SG) Tube lntearity

1. When the averaae reactor coolant svstem temperature is > 200°F the following shall be maintained:

A. SG Tube intearitv shall be maintained. and B. All SG tubes satisfvina the tube repair criteria shall be pluaaed in accordance with the Steam Generator Program.

Note: Separate entrv condition is allowed for each SG tube.

2. If the requirements of TS 3.1 .g.l .B are not met. then:

A. Within 7 davs verifv tube intearitv of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and B. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following the next refuelina outaae or SG tube inspection.

3. If the requirements of TS 3.1 .g.2.A or TS 3.1 .g.l .A are not met. then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

LAR 218

&dies to the surveillance reauirements for RCS operational LEAKAGE in TS 3.1 .d. I To assure that the RCS operational LEAKAGE reauirements are verified in a sufficient periodicitv.

Note 1: LE:AKAGE surveillances are not required to be ~erformed 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.

Note 2: T!; 4.18.a is not applicable to primary to secondarv LEAKAGE

- - I

a. Verifv RCS operational LEAKAGE, except for primary to secondary LEAKAGE. is within limits bv performance of RCS water inventory balance each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b. Verify primary to secondary LEAKAGE is < 150 aallons per dav through any one SG each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

LAR 21 8

b d i e s to the surveillance reauirements for Steam Generator (SG) Tube Integritv in TS To assure that the Steam Generator Tube lntegritv reauirements are verified in a sufficient periodicitv.

a. Verifv SG tube integritv in accordance with the Steam Generator Proaram.

b. Verifv that each inspected SG tube that satisfies the tube repair criteria is p l u g ~ e d in accordance with the Steam Generator Proaram prior to enterina INTERMEDIATE SHUTDOWN followina a SG tube inspection.

LAR 218

b. Whenever integrity of a pressure isolation valve listed in Table TS 3.1-2 cannot be demonstrated, the integrity of the remaining pressure isolation valve in each high pressure line having a leaking valve shall be determined and recorded daily.

In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.

b. Deleted-LAR 218

LAR 218 LAR 218 LAR 218 LAR 218 TABLE TS 4.2-2 STEAM GENERATOR TUBE INSPECTION IS been deleted LAR 218 Page 1 of 1

b. Unique Reporting Requirements

1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and 1V.C of Appendix I to 10 CFR Part 50.

2. Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operation of the unit for the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.l of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

(1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of lnspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

4. Steam Generator Tube lnspection RePort A report shall be submitted within 180 davs after the initial entrv into INTERMEDIATE SHUTDOWN followina completion of an inspection performed in accordance with the S~ecification6.22. Steam Generator

[SG) Program. The report shall include:

a. The scope of inspections Performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniaues utilized for each dearadation mechanism, LAR 21 8 TS 6.9-6

d. Location. orientation (if linear): and measured sizes (if available) of service induced indications, 9

dearadation mechanism,

f. Total number and percentaae of tubes pluaaed to date,

a. The results of condition monitorina. includina the results of tube pulls and in-situ testing,

h. The effective pluaaina percentaae for all pluggina in each SG.

LAR 21 8

6 . 2 2 iGFNFRBTOR !SG! PROGRAM I A Steam Generator Program shall be established and implemented to ensure that SG tube integritv is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitorina assessments. Condition monitorina assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural intearitv and accident induced leakaae. The "as found" condition refers to the condition of the tubina durina an SG inspection outaae, as determined from the inservice inspection results or by other means?prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage durinq which the SG tubes are inspected or pluaaed to confirm that the performance criteria are beina met.

b. Performance criteria for SG tube intearitv. SG tube integritv shall be maintained by meetina the performance criteria for tube structural intearitv. accident induced leakaae, and operational LEAKAGE.

.Structural intearitv performance criterion: All in-service steam aenerator tubes shall retain structural intearitv over the full range of normal operatina conditions (includina startup. operation in the power ranae. hot standby, and cool down and all anticipated transients included in the design specification) and desian basis accidents. This includes retainina a safetv factor of 3.0 aaainst burst under normal steadv state full power operation primary-to-secondary pressure differential and a safetv factor of 1.4 aaainst burst applied to the design basis accident primarv-to-secondary pressure differentials. Apart from the above requirements. additional loadina conditions associated with the design basis accidents, or combination of accidents in accordance with the desian and licensina basis. shall also be evaluated to determine if the associated loads contribute sianificantlv to burst or collapse. In the assessment of tube intearitv, those loads that do significantlv affect burst or colla~seshall be determined and assessed in combination with the loads due to pressure with a safetv factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage ~erformancecriterion: The primary to secondary accident induced leakage rate for any desian basis accident. other than a SG tube rupture. shall not exceed the leakaae rate assumed in the accident analvsis in terms of total leakage rate for all SGs and leakaae rate for an individual SG.

Leakaae is not to exceed 150 apd per SG.

3. The operational LEAKAGE performance criterion is specified in TS 3.1 .d. "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceedina 40% of the nominal tube wall thickness shall be plugged.

LAR 21 8

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the obiective of detectina flaws of anv type ( e . ~ .volumetric

, flaws. axial and circumferential cracks) that mav be present alona the length of the tube. from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfv the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meetina the requirements of d.1, d.2. and d.3 below, the inspection scope. inspection methods, and inspection intervals shall be such as to ensure that SG tube intearitv is maintained until the next SG inspection. An assessment of degra~dationshall be performed to determine the tvpe and location of flaws to which the tubes mav be susceptible and. based on this assessment, to determine which inspection methods need to be emploved and at what locations.

1. Inspect 100°/~of the tubes in each SG durina the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108. 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to beain after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes bv the refuelina outaae nearest the midpoint of the period and the remaining 50% bv the refuelina outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages [whichever is less) without beina inspected.

3. If crack indications are found in anv SG tube, then the next inspection for each SG for the dearadation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refuelina outage (whichever is less). If definitive information. such as from examination of a pulled tube! diaanostic non-destructive testina. or enaineerina evaluation indicates that a crack-like indication is not associated with a crack(s). then the indication need not be treated as a crack.

e. Provisions for monitoring operational primarv to secondary LEAKAGE.

LAR 21 8 I

ATTACHMENT 3 LICENSE AMENDMENT REQUEST 218 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY PROPOSED TECHNICAL SPECIFICATION PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

TABLE OF CONTENTS TECHNICAL SPECIFICATIONS APPENDIX A Section .

Title Page Definitions ............................................................................................................... 1.0.1 l.O.a Quadrant-to-Average Power Tilt Ratio ......................................................... 1.0.1 l.O.b Safety limits................................................................................................ 1.0-1 1.0.c Limiting Safety System Settings .................................................................. 1.0-1 l.O.d Limiting Conditions for Operation................................................................. 1.0-1 l.O.e Operable - Operability ................................................................................1.0-1 1.0..f Operating .................................................................................................. 1.0-1 1.0. g Containment System Integrity......................................................................1.0-2 l.O.h Protective Instrumentation Logic .................................................................1.0-2 l.O.i Instrumentation Surveillance ....................................................................... 1.0-3 l.O.j Modes ....................................................................................................... 1.0-4 l.O.k Reactor Critical...........................................................................................1.0-4 1.0.1 Refueling Operation ................................................................................... 1.0-4 l.O.m Rated Power ............................................................................................... 1.0-4 l.O.n Reportable Event ........................................................................... ,. ...........1.0-4 1.0.0 Radiological Effluents ................................................................................. 1.0-5 1.0.p Dose Equivalent 1-131 ................................................................................ 1.0-6 l.O.q Core Operating Limits Report ...................................................................... 1.0-6 l.O.r Shutdown Margin ....................................................................................... 1.0-6 1.0s Immediately................................................................................................1.0-6 l.O.t Leakage ....................................................................................................1.0-7 (

2.0 Safety Limits and Limiting Safety System Settings ..................................................... 2.1 -1 2.1 Safety Limits. Reactor Core ......................................................................... 2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure........................................... 2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation .......................................................................................... 2.3-1 2.3.a Reactor Trip Settings ................................................................ 2.3-1 2.3.a.l Nuclear Flux ........................................................2.3-1 2.3.a.2 Pressurizer.......................................................... 2.3-1 2.3.a.3 Reactor Coolant Temperature ............................. 2.3-2 2.3.a.4 Reactor Coolant Flow ......................................... .2.3-3 2.3.a.5 Steam Generators .............................................. .2.3-3 2.3.a.6 Reactor Trip Interlocks ........................................ 2.3-4 2.3.a.7 Other Trips .......................................................... 2.3-4 3.0 Limiting Conditions for Operation.............................................................................. 3.0.1 3.1 Reactor Coolant System............................................................................. 3.1-1 3.1.a Operational Components...........................................................3.1-1 3.1 .a. 1 Reactor Coolant Pumps ...................................... 3.1-1 3.1 .a.2 Decay Heat Removal Capability...........................3.1-1 3.1 .a.3 Pressurizer Safety Valves .................................... 3.1-3 3.1.a.4 Pressure Isolation Valves ....................................3.1-4 3.1 .a.5 Pressurizer PORV and PORV Block Valves .........3.1-4 3.1 .a.6 Pressurizer Heaters.............................................3.1-5 3.1 .a.7 Reactor Coolant Vent System..............................3.1-5 3.1 .b Heatup & Cooldown Limit Curves for Normal Operation.............3.1-6 3.1 .c Maximum Coolant Activity .........................................................3.1-7 3.1.d RCS Operational Leakage ......................................................... 3.1-8 1 3.1 .e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration .............................................................. 3.1-9 3.1 .f Minimum Conditions for Criticality ............................................ 3.1 -10 3.1 .g Steam Generator (SG) Tube Integrity ...................................... 3.1-1 1 1

Section Title Page 3.2 Chemical and Volume Control System ............................................................ 3.2.1 3.3 Engineered Safety Features and Auxiliary Systems ........................................ 3.3.1 3.3.a Accumulators ........................................................................... 3.3.1 3.3.b Emergency Core Cooling System ..............................................3.3.2 3.3.c Containment Cooling Systems...................................................3.3.4 3.3.d Component Cooling System ...................................................... 3.3.6 3.3.e Service Water System............................................................... 3.3.7 3.4 Steam and Power Conversion System ......................................................... 3.4.1 3.4.a Main Steam Safety Valves ........................................................ 3.4.1 3.4.b Auxiliary Feedwater System ...................................................... 3.4.1 3.4.c Condensate Storage Tank ......................................................... 3.4.3 3.4.d Secondary Activity Limits.......................................................... .3.4.3 3.5 Instrumentation System .............................................................................. 3.5.1 3.6 Containment System .................................................................................. 3.6.1 3.7 Auxiliary Electrical Systems .........................................................................3.7.1 3.8 Refueling Operations..................................................................................3.8.1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits ................................................. 3.10-1 3.1O.a Shutdown Reactivity............................................................... .3.10.1 3.10.b Power Distribution Limits ......................................................... 3.10.1 3.10.c Quadrant Power Tilt Limits ..................................................... .3.10.4 3.1O.d Rod Insertion Limits ................................................................ 3.10.4 3.1O.e Rod Misalignment Limitations ................................................. .3.10.5 3.10.f Inoperable Rod Position Indicator Channels ............................3.10.5 3.10.9 Inoperable Rod Limitations ..................................................... .3.10.7 3.10 .h Rod Drop Time ....................................................................... .3.10.7 3.10.i Rod Position Deviation Monitor............................................... .3.10.7 3.10.j Quadrant Power Tilt Monitor.................................................... 3.10.7 3.1O.k Core Average Temperature .....................................................

3.10-7 3.10.1 Reactor Coolant System Pressure.......................................... .3.10.7 3.1O.m Reactor Coolant Flow ............................................................ .3.10.8 3.10.n DNBR Parameters.................................................................. .3.10.8 3.1 1 Core Surveillance Instrumentation ............................................................. 3.1 1-1 3.12 Control Room Post-Accident Recirculation System ................................... .3.12.1 3.14 Shock Suppressors (Snubbers) ................................................................. 3.14.1 4.0 Surveillance Requirements.......................................................................................4.0.1 4.1 Operational Safety Review ......................................................................... .4. 1.1 4.2 ASME Code Class In-service Inspection and Testing .................................. 4.2.1 4.2.a ASME Code Class 1 , 2 . 3 , and MC Components and Supports ..................................................................................

4.2-1 4.2.b Deleted ...................................................................................

.4.2-2 1 4.3 Deleted

Section .

Title Containment Tests 4.4.1 4.4.a Integrated Leak Rate Tests (Type A) ....................................... 4.4.1 4.4.b Local Leak Rate Tests (Type B and C) ...................................... 4.4-1 4.4.c Shield Building Ventilation System............................................. 4.4-1 4.4.d Auxiliary Building Special Ventilation System ............................. 4.4-3 4.4.e Containment Vacuum Breaker System ......................................4.4-3 4.4.f Containment Isolation Device Position Verification ............................................................................... 4.4.3 Emergency Core Cooling System and Containment Air Cooling System Tests ................................................................................ 4.5-1 4.5.a System Tests ........................................................................... 4.5.1 4.5.a.l Safety Injection System ....................................... 4.5-1 4.5.a.2 Containment Vessel Internal Spray System ................................................................ 4.5.1 4.5.a.3 Containment Fan Coil Units ................................. 4.5-2 4.5.b Component Tests ..................................................................... .4.5-2 4.5.b.l Pumps ............................................................... .4.5-2 4.5.b.2 Valves ................................................................ .4.5-2 Periodic Testing of Emergency Power System.............................................4.6-1 4.6.a Diesel Generators .....................................................................4.6-1 4.6.b Station Batteries ........................................................................ 4.6-2 Main Steam Isolation Valves ....................................................................... 4.7-1 Auxiliary Feedwater System ........................................................................ 4.8-1 Reactivity Anomalies .................................................................................. 4.9-1 Deleted Deleted Spent Fuel Pool Sweep System ................................................................ 4.1 2-1 Radioactive Materials Sources ................................................................. .4.13.1 Testing and Surveillance of Shock Suppressors (Snubbers) ......................4.14.1 Deleted Reactor Coolant Vent System Tests ................................................ ., .........4.1 6-1 Control Room Postaccident Recirculation System .................................... .4.17-1 RCS Operational Leakage ......................................................................... 4.18-1 Steam Generator (SG) Tube Integrity ........................................................ 4.19-1 5.0 Design Features ...................................................................................................... 5.1.1 5.1 Site ...........................................................................................................5.1.1 5.2 Containment ............................................................................................... 5.2.1 5.2.a Containment System ................................................................ .5.2.1 5.2.b Reactor Containment Vessel .................................................... 5.2-2 5.2.c Shield Building .......................................................................... 5.2-2 5.2.d Shield Building Ventilation System............................................. 5.2-2 5.2.e Auxiliary Building Special Ventilation Zone and Special Ventilation System ........................................................ 5.2.2 5.3 Reactor Core .............................................................................................. 5.3-1 5.3.a Fuel Assemblies ........................................................................5.3-1 5.3.b Control Rod Assemblies ............................................................5.3.1 5.4 Fuel Storage .............................................................................................. 5.4-1 5.4.a Criticality .................................................................................. 5.4-1 5.4.b Capacity ................................................................................... 5.4-1 5.4.c Canal Rack Storage .................................................................. 5.4-1 TS iii

Section .

Title Paqe 6.0 Administrative Controls ............................................................................................6.1 -1 6.1 Responsibility ............................................................................................. 6.1-1 6.2 Organization............................................................................................... 6.2-1 6.2.a Off-Site Staff ............................................................................ 6.2-1 6.2.b Facility Staff ............................................................................. 6.2-1 6.2.c Organizational Changes ............................................................ 6.2-1 6.3 Plant Staff Qualifications ............................................................................ 6.3-1 6.4 Training ..................................................................................................... 6.4-1 6.5 Deleted ...................................................................................... 6.51 - 6.5-6 6.6 Deleted .....................................................................................................

. . . .6.6-1 6.7 Safety L~mitV~olat~on .................................................................................. 6.7-1 6.8 Procedures ............................................................................................... .6.8-1 6.9 Reporting Requirements............................................................................. 6.9-1 6.9.a Routine Reports ........................................................................ 6.9-1 6.9.a.l Startup Report ..................................................... 6.9-1 6.9.a.2 Annual Reporting Requirements ..........................6.9-1 6.9.a.3 Monthly Operating Report.................................... 6.9-3 6.9.a.4 Core Operating Limits Report ............................ .6.9-3 6.9.b Unique Reporting Requirements................................................ 6.9-6 6.9.b.l Annual Radiological Environmental Monitoring Report ............................................... .6.9-6 6.9.b.2 Radioactive Effluent Release Report ...................6.9-6 6.9.b.3 Special Reports .................................................. .6.9-6 6.9.b.4 Steam Generator Tube Inspection Report ...........6.9-6 1 6.10 Record Retention ..................................................................................... 6.10-1 6.1 1 Radiation Protection Program.................................................................... 6.1 1-1 6.12 System Integrity........................................................................................ 6.12-1 6.13 High Radiation Area ................................................................................. 6.13-1 6.14 Deleted .................................................................................................. .6.14-1 6.15 Secondary Water Chemistry ...................................................................... 6.15-1 6.16 Radiological Effluents............................................................................... 6.16-1 6.1'7 Process Control Program (PCP) ................................................................ 6.17-1 6.18 Offsite Dose Calculation Manual (ODCM) .................................................. 6.18-1 6.19 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems .........................................................6.19-1 6.20 Containment Leakage Rate Testing Program ............................................6.20-1 6.21 Technical Specifications (TS) Bases Control Program ...............................6.21 -1 6.22 Steam Generator (SG) Program ................................................................ 6.22-1 1 718.0 Deleted

LIST OF TABLES TABLE -

TITLE 1.O-1 ................. Frequency Notations 3.1-1 ................. Deleted 3.1-2 ................. Reactor Coolant System Pressure Isolation Valves 3.5-1 ................. Engineered Safety Features Initiation Instrument Setting Limits 3.5-2 ................. Instrument Operation Conditions for Reactor Trip 3.5-3 ................. Emergency Cooling 3.5-4 ................. Instrument Operating Conditions for Isolation Functions 3.5-5 .................Instrument Operation Conditions for Safeguards Bus Power Supply Functions 3.5-6 ................. Accident Monitoring Instrumentation Operating Conditions for Indication 4.1-1 ................. Minimum Frequencies for Checks, Calibrations and Test of lnstrument Channels 4.1-2 ................. Minimum Frequencies for Sampling Tests 4.1-3 .................Minimum Frequencies for Equipment Tests 4.2-1 ................. Deleted 4.2-2 ................. Deleted 4.2-3 ................. Deleted

t. LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

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

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

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

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

c. Pressure Boundary Leakage LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

d. RCS Operational LEAKAGE

1. When the average RCS temperature is > 200°F, 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).

2. If the limits contained in TS 3.1 .d.l are exceeded for reasons other than pressure boundary LEAKAGE or primary-to-secondary LEAKAGE, then reduce the LEAKAGE to within their limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3. If the limits contained in TS 3.1 .d.l for pressure boundary or primary to secondary LEAKAGE are exceeded, or the time limit contained in TS 3.1.d.2 is exceeded, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4. When the reactor is critical and above 2% power, two reactor coolant leak detection I systems of different operating principles shall be in operation with one of the two systems sensitive to radioactivity. Either system may be out of operation for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provided at least one system is OPERABLE.

g. Steam Generator (SG) Tube Integrity

1. When the average reactor coolant system temperature is > 200°F the following shall be maintained:

A. SG Tube integrity shall be maintained, and B. All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

Note: Separate entry condition is allowed for each SG tube.

2. If the requirements of TS 3.1 .g.l .B are not met, then:

A. Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and B. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following the next refueling outage or SG tube inspection.

3. If the requirements of TS 3.1 .g.2.A or TS 3.1 .g.l .A are not met, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4.1 8 RCS Operational LEAKAGE 1 APPLICABILITY I Applies to the surveillance requirements for RCS operational LEAKAGE in TS 3.1 .d. I OBJECTIVE I To assure that the RCS operational LEAKAGE requirements are verified in a sufficient periodicity.

SPECIFICATION Note 1: LEAKAGE surveillances are 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.

Note 2: TS 4.1 8.a is not applicable to primary to secondary LEAKAGE

a. Verify RCS operational LEAKAGE, except for primary to secondary LEAKAGE, is within limits by performance of RCS water inventory balance each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b. Verify primary to secondary LEAKAGE is I 150 gallons per day through any one SG each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

4.19 Steam Generator (SG) Tube Integrity I APPLICABILITY I Applies to the surveillance requirements for Steam Generator (SG) Tube lntegrity in TS 3.1 .g.

OBJECTIVE I To assure that the Steam Generator Tube lntegrity requirements are verified in a sufficient periodicity.

SPECIFICATION I

a. Verify SG tube integrity in accordance with the Steam Generator Program.

b. Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following a SG tube inspection.

b. Whenever integrity of a pressure isolation valve listed in Table TS 3.1-2 cannot be demonstrated, the integrity of the remaining pressure isolation valve in each high pressure line having a leaking valve shall be determined and recorded daily.

In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.

b. Deleted I

TABLE TS 4.2-2 STEAM GENERATOR TUBE INSPECTION TS Table 4.2-2 has been deleted Page 1 of 1

b. Unique Reporting Requirements

1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and 1V.C of Appendix I to 10 CFR Part 50.

2. Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operation of the unit for the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.l of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

(1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

4. Steam Generator Tube lns~ectionReport A r e ~ o r tshall be submitted within 180 davs after the initial entrv into INTERMEDIATE SHUTDOWN followina completion of an inspection performed in accordance with the S~ecification6.22, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active dearadation mechanisms found,

c. Nondestructive examination techniaues utilized for each dearadation mechanism,

d. Location. orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active dearadation mechanism,

f. Total number and percentage of tubes pluaaed to date,

_a. The results of condition monitorina. including the results of tube pulls and in-situ testing,

h. The effective pluaaina percentage for all pluaaina in each SG.

6.22 STEAM GENERATOR (SG) PROGRAM I A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 150 gpd per SG.

3. The operational LEAKAGE performance criterion is specified in TS 3.1 .d, "RCS Operational LEAKAGE."

Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. lnspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. lnspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

ATTACHMENT 4 LICENSE AMENDMENT REQUEST 218 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY MARKED UP TECHNICAL SPECIFICATION BASES PAGES FOR INFORMATION ONLY KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

RCS Operational LEAKAGE (TS 3.1.d)(Io' Components that contain or transport the coolant to or from the reactor core make up the RCS.

Component joints are made by weldina. bolting, rolling. or pressure loadina. and valves isolate connectina systems from the RCS.

Durina plant life. the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE technical specification (TS)- is to limit svstem operation in the presence of LEAKAGE from these sources to amounts that do not compromise safetv. This TS requirement specifies the types and amounts of LEAKAGE.

(17)

KPS USAR. GDC Criterion 16 - "Monitoring Reactor Coolant Pressure Boundary. , states that 3,

means shall be provided for monitorina the reactor coolant pressure boundary to detect leakage.

USAR section 6.5 describes the capabilities of the leakage monitorina indication svstems.

The safetv sianificance of RCS LEAKAGE varies widelv depending on its source. rate, and duration. Therefore. detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quicklv separatina the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators. allowina them to take corrective action should a leak occur that is detrimental to the safetv of the facility and the public.

A limited amount of leakaae inside containment is expected from auxiliarv svstems that cannot be made 100% leaktight. Leakaae from these systems should be detected. located. and isolated from the containment atmosphere. if possible. to not interfere with RCS leakaae detection.

This TS e w k e w w k d e a l s with protection of the reactor coolant pressure boundary (RCPB) from dearadation and the core from inadequate coolina. in addition to preventina the accident analvses radiation release assumptions from being exceeded. The consequences of violating this TS +wphwn&include the possibility of a loss of coolant accident (LOCA).

APPLICABLE Safetv Analvsis

-Except for primary to secondary LEAKAGE. the safetv analvses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safetv analvses for LOCA: the amount of leakaae can affect the probabilitv of such an event. The safetv analvsis for an event resulting in steam discharae to the atmosphere assumes that primary to secondarv LEAKAGE from the steam penerators (SGs) is 150 aallons per dav per steam aenerator (18)('9m'. The TS to limit primary to secondary LEAKAGE throuah anv one SG to less than or equal to 150 gallons per dav is the conditions assumed in the safetv analvsis.

Primarv to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. T: s k x w - e x k n t , oOther accidents or transients involve secondary steam release to the atmosphere. such as a steam penerator tube rupture

[SGTR). locked RCP rotor. and control rod ejection. The primary to secondary

'I6' USAR Sections 6.5, 11.2.3, 14.2.4 Kewaunee Power Station Updated Safety Analysis Report (USAR), Section 1.8, Criteria 16.

! I 8 ' USAR Section 14.2.4. "Steam Generator Tube Rupture.

""USAR Section 14.1.8, Locked Rotor

'20'USARSection 14.2.5. Main Steam Line Break

'"' Westinghouse Calculation CN-CRA-00-70. Rod Election Accident LAR 218 TS B3.1-9

L E A K A G E h k g e contaminates the secondary fluid.

The radiological accident Fanalvsis (22) for SGTR assumes the contaminated secondary fluid is released to the environment from the ruptured and the intact steam aenerators. The release from the ruptured SG occurs until 30 minutes after the reactor trip and the release from the intact SG occurs until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor trip when RHR is placed in service. The 150 gpd per SG primary to secondary LEAKAGE safetv analvsis assumption is relativelv inconsequential.

The SLB is less limiting for site radiation releases. The safetv analvsis for the SLB accident assumes 150 gpd primary to secondary LEAKAGE throuah the affected generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 50.67 or the staff approved licensina basis (i.e.. a small fraction of these limits).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

APPLICABILITY When the RCS average temperature is > 200°F. the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In COLD SHUTDOWN and REFUELING MODESS-!UTC?CU4LPd.LEAKAGE limits are not required because the reactor coolant pressure is far lower. resulting in lower stresses and reduced potentials for LEAKAGE.

SPECIFICATIONS RCS operational LEAKAGE shall be limited to:

A. Pressure Boundary LEAKAGE No pressure boundarv LEAKAGE is allowed. beina indicative of material deterioration.

LEAKAGE of this tvpe is unacceptable as the leak itself could cause further deterioration, resultina in hiaher LEAKAGE. Violation of this TS ea&wxm&could result in continued dearadation of the RCPB. LEAKAGE Dast seals and aaskets is not pressure boundary LEAKAGE.

B. Unidentified LEAKAGE One aallon per minute ( a ~ mof) unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this TS could result in continued dearadation of the RCPB, if the LEAKAGE is from the pressure boundary.

Leakaae from the Reactor Coolant Svstem is collected in the containment or bv the other closed svstems. These closed svstems are: the Steam and Feedwater Svstem. the Waste Disposal Svstem and the Component Coolina Svstem. Assuming the existence of the

'") Westinghouse Calculation CN-CRA-99-36. Steam Generator Tube Rupture 1 LAR 218 TS B3.1-10

maximum allowable activity in the reactor coolant, the rate of 1 apm unidentified leakaae would not exceed the limits of 10 CFR Part 20. This is shown as follows:

If the reactor coolant activity is WE pCi1cc ( E = average beta plus gamma energy per disintegration in Mev) and 1 gpm ofeakaae i s ~ s s u m e dto be discharaed throuah the air ejector, or throuah the Component Coolina Svstem vent line. then the vearlv whole bodv dose resultina from this activitv at the SlTE BOUNDARY. usina an annual averaae WQ = 2.0 x 10" sec/m3. is 0.09 remlvr. compared with the 10 CFR Part 20 limits of 0.1 remlvr.

With the limitina reactor coolant activitv and assumina initiation of a 1 gpm leak from the Reactor Coolant Svstem to the Component Coolina Svstem, the radiation monitor in the component cooling pump inlet header would annunciate in the control room. Operators would then investigate the source of the leak and take actions necessaty to isolate it. Should the leak result in a continuous discharae to the atmosphere via the component cooling surge tank and waste holdup tank. the resultant dose rate at the SlTE BOUNDARY would be 0.09 remlvr as aiven above.

Leakaae directly into the containment indicates the possibilitv of a breach in the coolant envelope. The limitation of 1 gpm for an unidentified source of leakaae is sufficientlv above the minimum detectable leak rate to provide a reliable indication of leakaae, and is well below the capacitv of one charaina pump (60 gpm).

C. ldentified 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 capabilitv of the RCS Makeup Svstem. ldentified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this TS could result in continued degradation of a component or svstem.

D. Primary to Secondary LEAKAGE throuah Anv One SG The limit of 150 aallons per dav limit per SG is based on the operational LEAKAGE performance criterion in NEI 97-06. Steam Generator Proaram Guidelines ('j'. The Steam Generator Program operational LEAKAGE performance criteria in NEI 97-06 states. "The RCS operational primary to secondary leakaae through anv one SG shall be limited to 150 aallons per dav." The limit is based on operating experience with SG tube degradation mechanisms that resulted in tube leakaae. The o~erationalleakage rate criterion in coniunction with the implementation of the Steam Generator Proaram is an effective measure for minimizina the frequency of steam generator tube ruptures.:

'"' NEI 97-06. "Steam Generator Program Guidelines." I LAR 218 TS B3.1-11

Unidentified LEAKAGE. or identified LEAKAGE:- LEA!#@ in excess of the TS wwwmeM-limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time allows time to verify leakaae rates and either identifv unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessarv to prevent further deterioration of the RCPB.

If any pressure boundary LEAKAGE exists. or primary to secondan, LEAKAGE is not within limits. or if unidentified or identified LEAKAGE;- )nE,MAG cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. the reactor must be brouaht to lower pressure conditions to reduce the severitv of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundarv LEAKAGE. The reactor must be brought to HOT SHUTDOWNlvGQH within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWNMQB&

within an additional 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />sMews. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary:.

The allowed Comdetion Times are reasonable. based on operating experience. to reach the required plant conditions from full power conditions in an orderlv manner and without challenging plant svstems. In COLD SHUTDOWNlvlWl3. the pressure stresses actina on the RCPB are much lower. and further deterioration is much less likelv.

1n rca D

7 wA 1 rn LAR 218 TS 63.1-12

LAR 218 If leakage is to the containment, it may be identified by one or more of the following methods:

A. The containment air particulate monitor is sensitive to low leak rates. The rates of reactor coolant leakage to which the instrument is sensitive are dependent upon the presence of corrosion product activity.

B. The containment radiogas monitor is less sensitive and is used as a backup to the air particulate monitor. The sensitivity range of the instrument is approximately 2 gpm to

> 10 gpm.

C. Humidity detection provides a backup to A: and B. The sensitivity range of the instrumentation I is from approximately 2 gpm to 10 gpm.

D. A leakage detection system is provided which determines leakage losses from all water and steam systems within the containment. This system collects and measures moisture condensed from the containment atmosphere by fancoils of the Containment Air Cooling System and thus provides a dependable and accurate means of measuring integrated total leakage, including leaks from the cooling coils themselves which are part of the containment boundary. The fancoil units drain to the containment sump, and all leakage collected by the containment sump will be pumped to the waste holdup tank. Pump running time will be monitored in the control room to indicate the quantity of leakage accumulated.

If leakage is to another closed system it will be detected by the area and process radiation monitors and/or inventory control.

LAR 218

In the event that the limits as provided in the COLR are not met, administrative rod withdrawal limits shall be developed to prevent further increases in temperature with a moderator temperature coefficient that is outside analyzed conditions. In this case, the calculated HFP moderator temperature coefficient will be made less negative by the same amount the hot zero power moderator temperature coefficient exceeded the limit as provided in the COLR. This will be accomplished by developing and implementing administrative control rod withdrawal limits to achieve a moderator temperature coefficient within the limits for HFP moderator temperature coefficient.

Due to the control rod insertion limits of TS 3.10.d and potentially developed control rod withdrawal limits, it is possible to have a band for control rod location at a given power level. The withdrawal limits are not required if TS 3.1 .f.3 is satisfied or if the reactor is subcritical.

If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits as provided in the COLR are not developed, then the plant shall be taken to HOT STANDBY until the moderator temperature coefficient is within the limits as specified in the COLR.

The reactor is allowed to return to criticality whenever TS 3.1.f is satisfied.

BASIS - Steam Generator Tube lntearlty [TS 3.1 .a BACKGROUND Steam aenerator (SG) tubes are small diameter. thin walled tubes that carry primary coolant throuah the primary to secondarv heat exchanaers. The SG tubes have a number of important safetv functions. Steam generator tubes are an intearal part of the reactor coolant pressure boundary (RCPB) and. as such. are relied on to maintain the primary svstem's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary svstem. In addition. as part of the RCPB. the SG tubes are unique in that thev act as the heat transfer surface between the primarv and secondary svstems to remove heat from the primarv svstem. This Specification addresses onlv the RCPB intearitv function of the SG. The SG heat removal function is addressed bv TS 3.4. "Steam and Power Conversion" when the RCS averaae temperature is areater than 350 F." and TS 3.1 .a.2. "Decav Heat Removal Capabilitv."

when the RCS temperature is less than or equal to 350 F.

SG tube intearitv means that the tubes are caoable of performina their intended RCPB safety function consistent with the licensina basis. including applicable regulatory requirements.

Steam aenerator tubing is subject to a varietv of degradation mechanisms. Steam generator tubes mav experience tube dearadation related to corrosion phenomena. such as wastaae, pitting. interaranular attack, and stress corrosion crackina. along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectivelv. The SG performance criteria are used to manage SG tube degradation.

Specification 6.22. "Steam Generator (SG) Program." requires that a program be established and implemented to ensure that SG tube intearitv is maintained. Pursuant to Specification 6.22. tube intearitv is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integritv. accident induced leakage. and operational LEAKAGE.

The SG performance criteria are described in Specification 6.22. Meeting the SG performance criteria provides reasonable assurance of maintainina tube integritv at normal and accident conditions.

LAR 218

The processes used to meet the SG performance criteria are defined bv the Steam Generator Proaram Guidelines.

APPLICABLE SAFETY ANALYSIS The steam generator tube rupture (SGTR) accident is the limiting desian basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analvsis of a SGTR event assumes a bounding ~rimarvto secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in TS 3.1 .d. "RCS Operational LEAKAGE," plus the leakaae rate associated with a double-ended rupture of a single tube.

The analvsis for desian basis accidents and transients other than a SGTR assume the SG tubes retain their structural intearitv (i.e.. thev are assumed not to rupture.) In these analvses, the steam discharae to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 300 aallons per dav. or is assumed to increase to 300 aallons per dav as a result of accident induced conditions. For accidents that do not involve fuel damaae. the primarv coolant activitv level of DOSE EQUIVALENT 1-131 is assumed to be equal to the TS 3.1 .c. "Maximum Coolant Activity." limits. For accidents that assume fuel damage. the primary coolant activitv is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of 10 CFR 50.67 or the NRC approved licensina basis (e.g.. a small fraction of these limits).

Steam generator tube intearitv satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

SPECIFICATIONS The TS requires that SG tube integritv be maintained. The TS also requires that all SG tubes that satisfv the repair criteria be pluqaed in accordance with the Steam Generator Proaram.

During an SG inspection. anv inspected tube that satisfies the Steam Generator Proaram repair criteria is removed from service by plugaina. If a tube was determined to satisfy the repair criteria but was not pluaaed. the tube mav still have tube intearitv.

In the context of this Specification. a SG tube is defined as the entire lenath of the tube, including the tube wall. between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered Dart of the tube.

A SG tube has tube intearitv when it satisfies the SG performance criteria. The SG performance criteria are defined in S~ecification6.22, "Steam Generator Proaram." and describe acceptable SG tube performance. The Steam Generator Proaram also provides the evaluation process for determinina conformance with the SG performance criteria.

There are three SG ~erformancecriteria: structural intearity. accident induced leakaae. and operational LEAKAGE. Failure to meet anv one of these criteria is considered failure to meet the TS.

The structural intearitv performance criterion provides a margin of safetv aaainst tube burst or collapse under normal and accident conditions. and ensures structural intearitv of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The aross structural failure of the tube wall. The condition tv~icallvcorresponds to an unstable openina displacement (e.g.. openina area increased in response to constant pressure)

LAR 218 TS B3.1-18

accompanied bv ductile (plastic) tearina of the tube material at the ends of the dearadation."

Tube collapse is defined as. "For the load displacement curve for a aiven structure, collapse occurs at the to^ of the load versus displacement curve where the slope of the curve becomes zero." The structural integritv ~erformancecriterion provides auidance on assessing loads that have a sianificant effect on burst or collapse. In that context. the term "sianificant" is defined as "An accident loading condition other than differential pressure is considered sianificant when the addition of such loads in the assessment of the structural integritv performance criterion could cause a lower structural limit or limitina burst/collapse condition to be established." For tube intearitv evaluations. except for circumferential degradation. axial thermal loads are classified as secondarv loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-bv-case basis. The division between primary and secondary classifications will be based on detailed analvsis and/or testina.

Structural intearitv requires that the primary membrane stress intensitv in a tube not exceed the yield strenath for all ASME Code. Section Ill. Service Level A (normal operatina conditions) and Service Level B (upset or abnormal conditions) transients included in the desian specification.

This includes safetv factors and applicable desian basis loads based on ASME Code. Section Ill, Subsection NB and Draft Reaulatorv Guide 1.121.

The accident induced leakage performance criterion ensures that the primarv to secondary LEAKAGE caused bv a desian basis accident. other than a SGTR, is within the accident analvsis assumptions. The accident analvsis assumes that accident induced leakaae does not exceed 150 aallons per day per SG. The accident induced leakaae rate includes anv primarv to secondarv LEAKAGE existina prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions durina plant operation. The limit on operational LEAKAGE is contained in TS 3.1 .d, "RCS Operational LEAKAGE." and limits primary to secondary LEAKAGE throuah anv one SG to 150 aallons per dav. This limit is based on the assumption that a single crack leaking this amount would not propaaate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack. the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam aenerator tube intearitv is challenaed when the Dressure differential across the tubes is large. Large differential pressures across SG tubes can onlv be experienced in the OPERATING. HOT STANDBY. HOT SHUTDOWN. or INTERMEDIATE SHUTDOWN MODES.

RCS conditions are far less challenging in the COLD SHUTDOWN or REFUELING MODES than durina the OPERATING. HOT STANDBY. HOT SHUTDOWN, or INTERMEDIATE SHUTDOWN MODES. In the COLD SHUTDOWN or REFUELING MODES, primary to secondary differential pressure is low. resulting in lower stresses and reduced ~otentialfor LEAKAGE, ACTIONS The ACTIONS are modified bv a Note clarifvina that the Conditions mav be entered inde~endentlvfor each SG tube. This is acceptable because the Reauired Actions provide appropriate compensatorv actions for each affected SG tube. Complvina with the Reauired Actions mav allow for continued operation. and subseauent affected SG tubes are aoverned by subseauent Condition entry and application of associated Required Actions.

LAR 218 TS B3.1-19

This TS applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfv the tube repair criteria but were not pluaaed in accordance with the Steam Generator Proaram as required bv TS 4.1 9. An evaluation of SG tube integritv of the affected tube(s) must be made. Steam aenerator tube integritv is based on meetina the SG performance criteria described in the Steam Generator Proaram. The SG repair criteria define limits on SG tube dearadation that allow for flaw growth between inspections while still providina assurance that the SG ~erformancecriteria will continue to be met. In order to determine if a SG tube that should have been ~ l u g p e dhas tube intearitv. an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outaae or SG tube insoection. The tube integritv determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated arowth of the dearadation prior to the next SG tube inspection. If it is determined that tube integritv is not being maintained. TS 3.1 .a.3 applies.

A Completion Time of 7 davs is sufficient to complete the evaluation while minimizina the risk of plant operation with a SG tube that mav not have tube intearitv.

If the evaluation determines that the affected tube(s) have tube integritv. Reauired Action TS 3.1.a.2.B allows plant operation to continue until the next refueling outaae or SG inspection provided the inspection interval continues to be supported bv an operational assessment that reflects the affected tubes. However. the affected tube(s) must be pluaaed prior to entering INTERMEDIATE SHUTDOWN followina the next refuelina outaae or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported bv the operational assessment.

If the Reauired Actions and associated Completion Times &are not met or if SG tube integritv is not beina maintained. the reactor must be brouqht to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The allowed Completion Times are reasonable, based on operatina experience. to reach the desired plant conditions from full power conditions in an orderlv manner and without challenaing plant svstems.

LAR 218

Verifvina RCS LEAKAGE to be within the TS ensures the intearitv of the RCPB is maintained.

Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can onlv be positivelv identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined bv performance of an RCS water inventory balance.

The RCS water inventory balance must be met with the reactor at steadv state operating conditions (stable temDerature. power level. pressurizer and makeup tank levels. makeup and letdown). This surveillance is modified bv 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 establishina steadv state operation. The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steadv state operation is required to perform a proper inventorv balance since calculations during maneuverina are not useful. For RCS operational LEAKAGE determination bv water inventorv balance. steadv state is defined as stable RCS pressure. temperature. power level, pressurizer and makeup tank levels. makeup and letdown. and RCP seal iniection and return flows.

An earlv warnina of pressure boundarv LEAKAGE or unidentified LEAKAGE is provided bv the automatic svstems that monitor the containment atmosphere radioactivitv and the containment sump level. It should be noted that LEAKAGE past seals and aaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in TS 3.1 .d.4.

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 aallons per dav cannot be measured accuratelv bv an RCS water inventory.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequencv is a reasonable interval to trend LEAKAGE and recoanizes the importance of earlv leakage detection in the prevention of accidents.

This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 aallons per day throuah anv one SG. Satisfvina the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Proaram is met. If this SR is h o t met. compliance with TS 3.1 .a. "Steam Generator (SG) Tube Intearitv."' should be evaluated. The 150 aallons per dav limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE throuah anv one SG.

If it is not practical to assian the LEAKAGE to an individual SG. all the primary to secondary LEAKAGE should be conservativelv assumed to be from one SG.

The surveillance is modified bv 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 steadv state operation. For RCS ~ r i m a r yto secondarv LEAKAGE determination. steadv state is defined as stable RCS Dressure, temperature, power level, wressurizer and makeup tank levels, makeup and letdown. and RCP seal injection and return flows.

LAR 218

The surveillance freauencv of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 farab samples in accordance with the EPRI guidelines"'.

'I' EPRI. " Pressurized Water Reactor Primary to Secondarv Leak Guidelines" I LAR 21 8 I

BASIS - Steam Generator (SG) Tube lntearity (TS 4.1 9)

Durina shutdown periods the SGs are inspected as required bv this SR and the Steam Generator Proaram. NEI 97-06, Steam Generator Proaram Guidelines. and its referenced EPRI Guidelines. establish the content of the Steam Generator Program.

Use of the Steam Generator Proaram ensures that the inspection is appropriate and consistent with accepted industry practices.

Durina SG inspections a condition monitorina assessment of the SG tubes is performed.

The condition monitorina assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitorina assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Proaram determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfving the tube repair criteria.

Inspection scope (i.e., which tubes or areas of tubina within the SG are to be inspected) is a function of existina and potential dearadation locations. The Steam Generator Proaram also specifies the inspection methods to be used to find potential dearadation.

Inspection methods are a function of degradation morphologv. non-destructive examination (NDE) technique capabilities. and inspection locations.

The Steam Generator Proaram defines the Freauencv of TS 4.19.a. The Freauencv is determined bv the operational assessment and other limits in the SG examination auidelines ( . The Steam Generator Program uses information on existina dearadations and arowth rates to determine an inspection Freauencv that provides reasonable assurance that the tubina will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.22 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

Durina an SG inspection. anv inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by pluggina. The tube repair criteria delineated in Specification 6.22 are intended to ensure that tubes accepted for continued service satisfv the SG performance criteria with allowance for error in the flaw size measurement and for future flaw arowth. In addition, the tube repair criteria. in coniunction with other elements of the Steam Generator Proaram. ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). NEI 97-06. "Steam Generator Program Guidelines." provides guidance for

(') EPRI. "Pressurized Water Reactor Steam Generator Examination Guidelines." 1 LAR 218 I

performina operational assessments to verifv that the tubes remainina in service will continue to meet the SG performance criteria.

The Freauencv of prior to enterina INTERMEDIATE SHUTDOWN following a SG inspection ensures that the Surveillance has been completed and all tubes meetina the repair criteria are pluaaed prior to subiectina the SG tubes to sianificant ~ r i m a r yto secondaty pressure differential.

LAR 218

Kewaunee Power Station (KPS) design was not designed to Section XI of the ASME Code; therefore, 100% compliance may not be practically achievable. However, the design process did consider access for in-service inspection, and made modifications within design limitations to provide maximum access. To the extent practical, Dominion Energy Kewaunee, Inc. performs inspection of ASME Code Class 1, Class 2, Class 3, and Class MC components in accordance with Section XI of the ASME Code. If an inspection required by the Code is impractical, Dominion Energy Kewaunee, Inc. requests Commission approval for deviation from the requirement.

The basis for surveillance testing of the Reactor Coolant System pressure isolation valves identified in Table TS 3.1-2 is contained within "Order for Modification of License" dated April 20, 1981.

Technical Specification 4.2.b (Deleted)

LAR 218

Technical Specification 4.2.b.l (Deleted)

Technical Specification 4.2.b.2 (Deleted)

Technical Specification 4.2.b.3 (Deleted) I LAR 218

Technical Specification 4.2.b.4 (Deleted)

Technical Specification 4.2.b.5 (Deleted)

Technical Specification 4.2.b.6 (Deleted)

Technical Specification 4.2.b.7 (Deleted]

tq%/\~ 4EQqF VVLV, "Dr.

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LAR 218

ATTACHMENT 5 LICENSE AMENDMENT REQUEST 218 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY PROPOSED TECHNICAL SPECIFICATION BASES PAGES FOR INFORMATION ONLY KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

RCS Operational LEAKAGE (TS 3.1.dlt'"'

Components that contain or transport the coolant to or from the reactor core make up the RCS.

Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE technical specification (TS) is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This TS requirement specifies the types and amounts of LEAKAGE.

KPS USAR, GDC Criterion 16 - "Monitoring Reactor Coolant Pressure Boundary, (17) , states that

?I means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage.

USAR section 6.5 describes the capabilities of the leakage monitoring indication systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This TS deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this TS include the possibility of a loss of coolant accident (LOCA).

APPLICABLE Safety Analysis Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to seconda LEAKAGE from the steam generators (SGs) is 150 gallons per day per steam generator (18)(1320)(21). The TS to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is the conditions assumed in the safety analysis.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. Other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). locked RCP rotor, and control rod ejection. The primary to secondary LEAKAGE contaminates the secondary fluid.

'I6) USAR Sections 6.5, 11.2.3, 14.2.4

'I7' Kewaunee Power Station Updated Safety Analysis Report (USAR), Section 1.8, Criteria 16.

( I X 'USAR Section 14.2.4, "Steam Generator Tube Rupture.

""USAR Section 14.1.8, Locked Rotor

'20'USARSection 14.2.5, Main Steam Line Break

"" Westinghouse Calculation CN-CRA-00-70, Rod Ejection Accident

The radiological accident analysis (22) for SGTR assumes the contaminated secondary fluid is released to the environment from the ruptured and the intact steam generators. The release from the ruptured SG occurs until 30 minutes after the reactor trip and the release from the intact SG occurs until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor trip when RHR is placed in service. The 150 gpd per SG primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

The SLB is less limiting for site radiation releases. The safety analysis for the SLB accident assumes 150 gpd primary to secondary LEAKAGE through the affected generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 50.67 or the staff approved licensing basis (i.e., a small fraction of these limits).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

APPLICABILITY When the RCS average temperature is > 200°F, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In COLD SHUTDOWN and REFUELING MODES, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

SPECIFICATIONS RCS operational LEAKAGE shall be limited to:

A. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this TS 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 containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this TS could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

Leakage from the Reactor Coolant System is collected in the containment or by the other closed systems. These closed systems are: the Steam and Feedwater System, the Waste Disposal System and the Component Cooling System. Assuming the existence of the maximum allowable activity in the reactor coolant, the rate of 1 gpm unidentified leakage would not exceed the limits of 10 CFR Part 20. This is shown as follows:

'"'Westinghouse Calculation CN-CRA-99-36, Steam Generator Tube Rupture 1 TS B3.1-10

(E If the reactor coolant activity is 911E ~ C i l c c = average beta plus gamma energy per disintegration in Mev) and 1 gpm of leakage is assumed to be discharged through the air ejector, or through the Component Cooling System vent line, then the yearly whole body dose resulting from this activity at the SlTE BOUNDARY, using an annual average WQ = 2.0 x sec/m3, is 0.09 remlyr, compared with the 10 CFR Part 20 limits of 0.1 remlyr.

With the limiting reactor coolant activity and assuming initiation of a 1 gpm leak from the Reactor Coolant System to the Component Cooling System, the radiation monitor in the component cooling pump inlet header would annunciate in the control room. Operators would then investigate the source of the leak and take actions necessary to isolate it. Should the leak result in a continuous discharge to the atmosphere via the component cooling surge tank and waste holdup tank, the resultant dose rate at the SlTE BOUNDARY would be 0.09 remlyr as given above.

Leakage directly into the containment indicates the possibility of a breach in the coolant envelope. The limitation of 1 gpm for an unidentified source of leakage is sufficiently above the minimum detectable leak rate to provide a reliable indication of leakage, and is well below the capacity of one charging pump (60 gpm).

C. ldentified 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 containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this TS 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 limit per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (23'. The Steam Generator Program operational LEAKAGE performance criteria 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 resulted 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.

Unidentified LEAKAGE, or identified LEAKAGE in excess of the TS limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-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 (3) NEl 97-06, "Steam Generator Program Guidelines."

TS B3.1-11

shut down. This action is necessary to prevent further deterioration of the RCPB.

If any pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limits, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. The reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

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. In COLD SHUTDOWN, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

If leakage is to the containment, it may be identified by one or more of the following methods:

The containment air particulate monitor is sensitive to low leak rates. The rates of reactor coolant leakage to which the instrument is sensitive are dependent upon the presence of corrosion product activity.

The containment radiogas monitor is less sensitive and is used as a backup to the air particulate monitor. The sensitivity range of the instrument is approximately 2 gpm to

> 10 gpm.

Humidity detection provides a backup to A and B. The sensitivity range of the instrumentation I is from approximately 2 gpm to 10 gpm.

A leakage detection system is provided which determines leakage losses from all water and steam systems within the containment. This system collects and measures moisture condensed from the containment atmosphere by fancoils of the Containment Air Cooling System and thus provides a dependable and accurate means of measuring integrated total leakage, including leaks from the cooling coils themselves which are part of the containment boundary. The fancoil units drain to the containment sump, and all leakage collected by the containment sump will be pumped to the waste holdup tank. Pump running time will be monitored in the control room to indicate the quantity of leakage accumulated.

If leakage is to another closed system it will be detected by the area and process radiation monitors and/or inventory control.

In the event that the limits as provided in the COLR are not met, administrative rod withdrawal limits shall be developed to prevent further increases in temperature with a moderator temperature coefficient that is outside analyzed conditions. In this case, the calculated HFP moderator temperature coefficient will be made less negative by the same amount the hot zero power moderator temperature coefficient exceeded the limit as provided in the COLR. This will be accomplished by developing and implementing administrative control rod withdrawal limits to achieve a moderator temperature coefficient within the limits for HFP moderator temperature coefficient.

Due to the control rod insertion limits of TS 3.10.d and potentially developed control rod withdrawal limits, it is possible to have a band for control rod location at a given power level. The withdrawal limits are not required if TS 3.1 .f.3 is satisfied or if the reactor is subcritical.

If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits as provided in the COLR are not developed, then the plant shall be taken to HOT STANDBY until the moderator temperature coefficient is within the limits as specified in the COLR.

The reactor is allowed to return to criticality whenever TS 3.1 .f is satisfied.

BASIS - Steam Generator Tube lnteqrity (TS 3.1 .a BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by TS 3.4. "Steam and Power Conversion" when the RCS average temperature is greater than 350 F," and TS 3.1 .a.2, "Decay Heat Removal Capability,"

when the RCS temperature is less than or equal to 350 F.

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.22, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.22, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE.

The SG performance criteria are described in Specification 6.22. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines.

APPLICABLE SAFETY ANALYSIS The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in TS 3.1 .d, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 300 gallons per day, or is assumed to increase to 300 gallons per day as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the TS 3.1.c, "Maximum Coolant Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of 10 CFR 50.67 or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

SPECIFICATIONS The TS requires that SG tube integrity be maintained. The TS also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.22, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the TS.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure)

accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation."

Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code, Section Ill, Subsection NB and Draft Regulatory Guide 1.121.

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 150 gallons per day per SG. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in TS 3.1 .d, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in the OPERATING, HOT STANDBY, HOT SHUTDOWN, or INTERMEDIATE SHUTDOWN MODES.

RCS conditions are far less challenging in the COLD SHUTDOWN or REFUELING MODES than during the OPERATING, HOT STANDBY, HOT SHUTDOWN, or INTERMEDIATE SHUTDOWN MODES. In the COLD SHUTDOWN or REFUELING MODES, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

This TS applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by TS 4.19. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, TS 3.1.g.3 applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action TS 3.1.g.2.B allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering INTERMEDIATE SHUTDOWN following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

If the Required Actions and associated Completion Times are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

BASIS - RCS O~erationalLEAKAGE (TS 4.18) I Verifying RCS LEAKAGE to be within the TS ensures the integrity of the RCPB is maintained.

Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown). This 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. The 12-hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper 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 seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in TS 3.1 .d.4.

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.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

This SR verifies that primary to secondary LEAKAGE is less than 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 TS 3.1 .g, "Steam Generator (SG) Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. 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 of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 samples in accordance with the EPRI guidelines"'.

"' EPRI, " Pressurized Water Reactor Primary to Secondary Leak Guidelines" I

BASIS - Steam Generator (SG) Tube Integrity (TS 4.1 9)

During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines, and its referenced EPRI Guidelines, establish the content of the Steam Generator Program.

Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed.

The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria.

lnspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.

lnspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of TS 4.1 9.a. The Frequency is determined by the operational assessment and other limits in the SG examination guidelines ( I ) . The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.22 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 6.22 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). NEI 97-06, "Steam Generator Program Guidelines." provides guidance for (I' EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines." I TS B4.19-1 I

performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering INTERMEDIATE SHUTDOWN following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

BASIS Kewaunee Power Station (KPS) design was not designed to Section XI of the ASME Code; therefore, 100°/~compliance may not be practically achievable. However, the design process did consider access for in-service inspection, and made modifications within design limitations to provide maximum access. To the extent practical, Dominion Energy Kewaunee, Inc. performs inspection of ASME Code Class 1, Class 2, Class 3, and Class MC components in accordance with Section XI of the ASME Code. If an inspection required by the Code is impractical, Dominion Energy Kewaunee, Inc. requests Commission approval for deviation from the requirement.

The basis for surveillance testing of the Reactor Coolant System pressure isolation valves identified in Table TS 3.1-2 is contained within "Order for Modification of License" dated April 20, 1981.

Technical Specification 4.2.b (Deleted)

Technical Specification 4.2.b.l (Deleted)

Technical Specification 4.2.b.2 (Deleted)

Technical Specification 4.2.b.3 (Deleted)

Technical Specification 4.2.b.4 {Deleted)

Technical Specification 4.2.b.5 (Deleted)

Technical Specification 4.2.b.6 (Deleted)

Technical Specification 4.2.b.7 (Deleted)