L-PI-25-026, Application to Revise Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant Leakage Requirements

Application to Revise Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant Leakage Requirements

ML25247A026
Person / Time
Site:	Prairie Island
Issue date:	09/03/2025
From:	Paulhardt W Northern States Power Company, Minnesota, Xcel Energy Inc
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-25-026
Download: ML25247A026 (1)

Text

1717 Wakonade Drive Welch, MN 55089 Connected

• Committed

• Trustworthy

• Safe L-PI-25-026 10 CFR 50.90 September 3, 2025 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 Application to Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), is submitting a request for an amendment to the Technical Specifications (TS) for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2.

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

which is an approved change to the Standard Technical Specifications (STS), into the PINGP, Units 1 and 2 TS. TSTF-554 revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected and adds a Required Action when pressure boundary leakage is identified.

The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS pages. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only.

NSPM requests that the amendment be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within six months of completion of the NRC's acceptance review. Once approved, the amendment shall be implemented within 120 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Minnesota State Official.

(}, Xcel Energy*

Document Control Desk L-Pl-25-026 Page 2 If you should have any questions regarding this submittal, please contact Ron Jacobson at (612) 330-6542 or ronald.g.jacobson@xcelenergy.com.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury, that the foregoing is true and correct.

Exec,te on Se tember J_, 2025.

\\ --

Werner K. Paulh rd Jr.

Site Vice Preside rairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region 111, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

L-PI-25-026 NSPM Enclosure Page 1 of 5 License Amendment Request Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements"

1.0 DESCRIPTION

In accordance with the provisions of 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), hereby requests adoption of TSTF-554, Revision 1, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the Prairie Island Nuclear Generating Plant (PINGP), Unit 1 and Unit 2, Technical Specifications (TS). The proposed amendment revises the TS definition of "Leakage" and the Reactor Coolant System (RCS) Operational Leakage TS to clarify the requirements.

2.0 ASSESSMENT

2.1 Applicability of Safety Evaluation NSPM has reviewed the safety evaluation for TSTF-554, Revision 1, provided to the Technical Specifications Task Force (TSTF) in a letter dated December 18, 2020, Agencywide Documents Access and Management System (ADAMS) Accession Nos.

ML20168A361 and ML20322A024. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-554, Revision 1. As described herein, NSPM has concluded that the justifications presented in TSTF-554, Revision 1, and the safety evaluation prepared by the NRC staff are applicable to the PINGP and justify this amendment for the incorporation of the changes into the PINGP TS.

2.2 Variations NSPM is proposing the following editorial variations from the TS changes described in TSTF-554, Revision 1. The existing PINGP TS differences as described in sections c.

through e. below follow from PINGP license amendments 91/84 (dated October 27, 1989, ML022210226), and license amendments 158/149 when the PINGP Improved Technical Specifications (ITS) conversion occurred (dated July 27, 2002, ML022070613).

a.

The PINGP TS 1.1 Definition for "Identified LEAKAGE," Item a.3., already ends with a ";" so the markups in this license amendment request differ editorially from TSTF-554 which shows the punctuation changing from a "," to a ";". These differences do not affect the applicability of TSTF-554 to the PINGP TS.

b.

The PINGP TS use different numbering than the Standard Technical Specifications on which TSTF-554 was based. Specifically, the PINGP TS

L-PI-25-026 NSPM Enclosure Page 2 of 5 section for RCS Leakage is 3.4.14 rather than 3.4.13 in the approved TSTF-554 model. This difference is administrative and does not affect the applicability of TSTF-554 to the PINGP TS.

c.

PINGP TS 3.4.14 contains four Conditions that encompass the two Conditions of the STS.

1.

For the original STS Condition A, the corresponding existing PINGP TS 3.4.14 Conditions are A and C. These two Conditions use "unidentified LEAKAGE" and "identified LEAKAGE" and together meet the "operational LEAKAGEfor reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE" of the original STS 3.4.13 Condition A.

2.

For the original STS Condition B, the corresponding existing PINGP TS 3.4.14 Conditions are B and D.

These differences do not affect the applicability of TSTF-554 to the PINGP TS.

d.

To retain the time allowances authorized by TSTF-554, The PINGP TS will have one additional Condition added. The rationale for this addition is as follows:

1.

While "Pressure boundary LEAKAGE exists." is broken out from PINGP TS D, the new Condition A for the PINGP TS matches the TSTF-554.

2.

To retain the same Required Action and Completion Time from the PINGP TS Condition D and match the time allowance of TSTF-554, a new PINGP Condition B is added retaining the MODE 3 and MODE 5 Completion Times of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> respectively for the new Condition A.

3.

The original PINGP TS Conditions, Required Actions, and Completion Times are renumbered sequentially starting with new Condition C and continuing through D, E, and F.

These differences do not affect the applicability of TSTF-554 to the PINGP TS.

e.

The PINGP TS contain different Required Actions and Completion Times than the STS; however, TSTF-554 does not change Required Actions or Completion Times with the exception of the creation of the new Condition A Required Action and the associated Completion Time. Similarly the only Completion Time adjustment in the PINGP TS is the addition of the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of new Condition A.

These differences do not affect the applicability of TSTF-554 to the PINGP TS.

f.

The Traveler and safety evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). PINGP was not licensed to the 10 CFR 50, Appendix A, GDC. The PINGP was designed and constructed to comply with NSPs

L-PI-25-026 NSPM Enclosure Page 3 of 5 understanding of the intent of the AEC General Design Criteria for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Since the construction of the plant was significantly completed prior to the issuance of the February 20, 1971, 10CFR50, Appendix A GDC, the plant was not reanalyzed, and the Final Safety Analysis Report (FSAR) was not revised to reflect these later criteria.

However, the AEC Safety Evaluation Report acknowledged that the AEC staff assessed the plant, as described in the FSAR, against the Appendix A design criteria and "... are satisfied that the plant design generally conforms to the intent of these criteria." This difference does not affect the applicability of TSTF-554 to the PINGP TS.

g.

Additional format changes are being made to page 1.1-4 for consistency within TS Section 1.1. These editorial changes do not affect the applicability of TSTF-554 to the PINGP TS.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis In accordance with the provisions of 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), hereby requests adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements,"

Revision 1, that is an approved change to the Standard Technical Specifications (STS) into the Prairie Island Nuclear Generating Plant (PINGP), Unit 1 and Unit 2, Technical Specifications (TS). The proposed amendment revises the TS definition of "Leakage,"

clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified.

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

1.

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

Response: No The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified.

The proposed change revises the definition of pressure boundary leakage.

Pressure boundary leakage is a precursor to some accidents previously evaluated. The proposed change expands the definition of pressure boundary leakage by eliminating the qualification that pressure boundary leakage must be from a "nonisolable" flaw. A new TS Action is created which requires isolation of

L-PI-25-026 NSPM Enclosure Page 4 of 5 the pressure boundary flaw from the Reactor Coolant System. This new action provides assurance that the flaw will not result in any accident previously evaluated.

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

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

2.

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

Response: No The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected and adds a Required Action when pressure boundary leakage is identified. The proposed change does not alter the design function or operation of the RCS. The proposed change does not alter the ability of the RCS to perform its design function. Since pressure boundary leakage is an evaluated accident, the proposed change does not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

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

3.

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

Response: No The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected and adds a Required Action when pressure boundary leakage is identified. The proposed change does not affect the initial assumptions, margins, or controlling values used in any accident analysis. The amount of leakage allowed from the RCS is not increased.

The proposed change does not affect any design basis or safety limit or any Limiting Condition for Operation.

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

L-PI-25-026 NSPM Enclosure Page 5 of 5 Based on the above, NSPM concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

ENCLOSURE ATTACHMENT 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AND UNIT 2 License Amendment Request Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" TECHNICAL SPECIFICATIONS PAGES (MARKUP)

(5 Pages Follow)

Definitions 1.1 Prairie Island Unit 1 - Amendment No. 158 177 206 Units 1 and 2 1.1-3 Unit 2 - Amendment No. 149 167 193 1.1 Definitions (continued)

DOSE DOSE EQUIVALENT XE-133 shall be that concentration of EQUIVALENT Xe-133 (microcuries per gram) that alone would produce the XE-133 same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

LEAKAGE LEAKAGE from the Reactor Coolant System (RCS) 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.

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; to and TBD TBD

Definitions 1.1 Prairie Island Unit 1 - Amendment No. 158 206 Units 1 and 2 1.1-4 Unit 2 - Amendment No. 149 193 1.1 Definitions

c.

Pressure Boundary LEAKAGE (continued)

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

MASTER A MASTER RELAY TEST shall consist of energizing all RELAY master relays in the channel required for channel OPERABILITY TEST and verifying the OPERABILITY of each required master relay.

The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE -

A system, subsystem, train, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS PHYSICS TESTS shall be those tests performed to measure the TESTS fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Appendix J of the USAR, Pre-Operational and Startup Tests;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

LEAKAGE (continued)

(continued)

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

TBD TBD-1

RCS Operational LEAKAGE 3.4.14 Prairie Island Unit 1 - Amendment No. 158 Units 1 and 2 3.4.14-1 Unit 2 - Amendment No. 149 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE LCO 3.4.14 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

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

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS unidentified LEAKAGE not within limit.

A.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Required Action and associated Completion Time of Condition A not met.

B.1 Be in MODE 3.

AND B.2.1 Identify LEAKAGE.

OR B.2.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 54 hours 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> INSERT 1 C

C D

D D

D C

TBD TBD

INSERT 1 A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B.

Required Action and associated Completion Time of Condition A not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours

RCS Operational LEAKAGE 3.4.14 Prairie Island Unit 1 - Amendment No. 158 177 Units 1 and 2 3.4.14-2 Unit 2 - Amendment No. 149 167 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS identified LEAKAGE not within limit for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

C.1 Be in MODE 3.

AND C.2.1 Reduce LEAKAGE to within limits.

OR C.2.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 14 hours 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> D. Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

D.1 Be in MODE 3.

AND D.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours E

TBD TBD E

E E

F F

F LJ I

~

I r

-~

-~

ENCLOSURE ATTACHMENT 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AND UNIT 2 License Amendment Request Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" TECHNICAL SPECIFICATIONS PAGES (RE-TYPED)

(8 Pages Follow)

Definitions 1.1 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 1.1-3 Unit 2 - Amendment No. TBD 1.1 Definitions (continued)

DOSE DOSE EQUIVALENT XE-133 shall be that concentration of EQUIVALENT Xe-133 (microcuries per gram) that alone would produce the XE-133 same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

LEAKAGE LEAKAGE from the Reactor Coolant System (RCS) 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 to not interfere with the operation of leakage detection systems; or

3.

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

Definitions 1.1 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 1.1-4 Unit 2 - Amendment No. TBD 1.1 Definitions (continued)

LEAKAGE

c.

Pressure Boundary LEAKAGE (continued)

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

MASTER A MASTER RELAY TEST shall consist of energizing all RELAY master relays in the channel required for channel OPERABILITY TEST and verifying the OPERABILITY of each required master relay.

The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE-A system, subsystem, train, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

Definitions 1.1 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 1.1-5 Unit 2 - Amendment No. TBD 1.1 Definitions (continued)

PHYSICS PHYSICS TESTS shall be those tests performed to measure the TESTS fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Appendix J of the USAR, Pre-Operational and Startup Tests;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that provides the reactor TEMPERATURE vessel pressure and temperature limits, including heatup and LIMITS cooldown rates, and the OPPS arming temperature for the current REPORT reactor vessel fluence period. These pressure and temperature limits (PTLR) shall be determined for each fluence period in accordance with Specification 5.6.6. Plant operation within these operating limits is addressed in LCO 3.4.3, RCS Pressure and Temperature (P/T)

Limits, LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -Reactor Coolant System Cold Leg Temperature (RCSCLT)

> Safety Injection (SI) Pump Disable Temperature, and LCO 3.4.13, Low Temperature Overpressure Protection (LTOP)

- Reactor Coolant System Cold Leg Temperature (RCSCLT)

< Safety Injection (SI) Pump Disable Temperature.

QUADRANT QPTR shall be the ratio of the maximum upper excore detector POWER TILT calibrated output to the average of the upper excore detector RATIO (QPTR) calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED RTP shall be a total reactor core heat transfer rate to the reactor THERMAL coolant of 1677 MWt.

POWER (RTP)

Definitions 1.1 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 1.1-6 Unit 2 - Amendment No. TBD 1.1 Definitions (continued)

REACTOR The RTS RESPONSE TIME shall be that time interval from when TRIP the monitored parameter exceeds its RTS trip setpoint at the channel SYSTEM (RTS) sensor output until opening of a reactor trip breaker. The response

RESPONSE

time may be measured by means of any series of sequential, TIME overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components have been evaluated in accordance with an NRC approved methodology.

SHUTDOWN SDM shall be the instantaneous amount of reactivity by which:

MARGIN (SDM)

a.

The reactor is subcritical; or

b.

The reactor would be subcritical from its present condition assuming all rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design temperature.

SLAVE A SLAVE RELAY TEST shall consist of energizing all slave relays RELAY in the channel required for channel OPERABILITY and verifying TEST the OPERABILITY of each required slave relay. The SLAVE shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlappping, or total steps.

THERMAL THERMAL POWER shall be the total reactor core heat transfer POWER rate to the reactor coolant.

Definitions 1.1 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 1.1-7 Unit 2 - Amendment No. TBD 1.1 Definitions (continued)

TRIP A TADOT shall consist of operating the trip actuating device and ACTUATING verifying the OPERABILITY of all devices in the channel DEVICE required for trip actuating device OPERABILITY. The TADOT OPERATIONAL shall include adjustment, as necessary, of the trip actuating device TEST so that it actuates at the required setpoint within the necessary (TADOT) accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

Definitions 1.1 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 1.1-8 Unit 2 - Amendment No. TBD Table 1.1-1 (page 1 of 1)

MODES MODE TITLE REACTIVITY CONDITION (keff)

% RATED THERMAL POWER(a)

AVERAGE REACTOR COOLANT TEMPERATURE

(°F) 1 Power Operation

> 0.99

> 5 NA 2

Startup

> 0.99

< 5 NA 3

Hot Standby

< 0.99 NA

> 350 4

Hot Shutdown(b)

< 0.99 NA 350 > Tavg > 200 5

Cold Shutdown(b)

< 0.99 NA

< 200 6

Refueling(c)

NA NA NA (a)

Excluding decay heat.

(b)

All reactor vessel head closure bolts fully tensioned.

(c)

One or more reactor vessel head closure bolts less than fully tensioned.

RCS Operational LEAKAGE 3.4.14 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 3.4.14-1 Unit 2 - Amendment No. TBD 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE LCO 3.4.14 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

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

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Required Action and associated Completion Time of Condition A not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours

RCS Operational LEAKAGE 3.4.14 Prairie Island Unit 1 - Amendment No. TBD Units 1 and 2 3.4.14-2 Unit 2 - Amendment No. TBD ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS unidentified LEAKAGE not within limit.

C.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. Required Action and associated Completion Time of Condition C not met.

D.1 Be in MODE 3.

AND D.2.1 Identify LEAKAGE.

OR D.2.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 54 hours 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> E. RCS identified LEAKAGE not within limit for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

E.1 Be in MODE 3.

AND E.2.1 Reduce LEAKAGE to within limits.

OR E.2.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 14 hours 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> F. Primary to secondary LEAKAGE not within limit.

F.1 Be in MODE 3.

AND F.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours

ENCLOSURE ATTACHMENT 3 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AND UNIT 2 License Amendment Request Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" TECHNICAL SPECIFICATIONS BASES PAGES (MARKUP)

(For Information Only)

(7 Pages Follow)

RCS Operational LEAKAGE B 3.4.14 Prairie Island Units 1 and 2 B 3.4.14-4 Revision 242 BASES APPLICABLE The USAR (Ref. 2) analysis for SGTR assumes the plant has been SAFETY operating with a primary to secondary leak rate for a period of time ANALYSES sufficient to establish radionuclide equilibrium in the secondary loop (continued) at the applicable limits.

The safety analysis for the SLB accident assumes the total primary to secondary LEAKAGE is 1 gallon per minute from the faulted SG or is assumed to increase to 1 gallon per minute as a result of accident induced conditions plus 150 gallons per day from the intact SG. The safety analysis for other secondary side accidents assume a steady state leakage of 150 gpd to the intact Steam Generator(s).

The safety analysis assumes the leakage from the faulted SG will be limited to 1 gpm. The dose consequences resulting from the secondary side accidents 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).

LCO 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 LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Seal welds are provided at the threaded joints of all reactor vessel head penetrations (spare penetrations, full-length Control Rod Drive Mechanisms, and thermocouple columns). Although these seals are part of the RCPB as defined in 10CFR50 Section 50.2, minor leakage past the seal weld is not a fault in the RCPB Pressure prohibited RCPB TBD

/4~ _______,

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RCS Operational LEAKAGE B 3.4.14 Prairie Island Units 1 and 2 B 3.4.14-5 Revision 242 BASES LCO or a structural integrity concern. Pressure retaining components (continued) are differentiated from leakage barriers in the ASME Boiler and Pressure Vessel Code. In all cases, the joint strength is provided by the threads of the closure joint.

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 LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified leakage must be evaluated to assure that continued operation is safe.

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 LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 3). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience TBD Separating the sources of LEAKAGE (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

RCS Operational LEAKAGE B 3.4.14 Prairie Island Units 1 and 2 B 3.4.14-6 Revision 242 BASES LCO with SG tube degradation mechanism that results in tube (continued) 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.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

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

LCO 3.4.15, RCS Pressure Isolation Valve (PIV) Leakage, measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS A.1 Unidentified LEAKAGE in excess of the LCO limits must be identified or 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 shut down. This action is necessary to prevent further deterioration of the RCPB.

B.1, B.2.1, and B.2.2 If unidentified LEAKAGE cannot be identified or 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 TBD C

D D

D INSERT BASES 1 IJ I

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

B.1 If the pressure boundary LEAKAGE isolation Required Action of Condition A cannot be met within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time, the reactor must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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

RCS Operational LEAKAGE B 3.4.14 Prairie Island Units 1 and 2 B 3.4.14-7 Revision 242 BASES ACTIONS B.1, B.2.1, and B.2.2 (continued)

(continued) pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals, gaskets, and pressurizer safety valves seats is not pressure boundary LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the LEAKAGE source cannot be identified within 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br />, then the reactor must be placed in MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-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 MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

C.1, C.2.1, and C.2.2 If RCS identified LEAKAGE, other than pressure boundary LEAKAGE or primary to secondary LEAKAGE, is not within limits, then the reactor must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In this condition, 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> are allowed to reduce the identified leakage to within limits. If the identified LEAKAGE is not within limits within this time, the reactor must be placed in MODE 5 within 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />.

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

D.1 and D.2 If RCS pressure boundary LEAKAGE exists or if primary to secondary LEAKAGE (150 gpd limit) is not within limits, the D

D D

TBD E

E E

F F

RCS Operational LEAKAGE B 3.4.14 Prairie Island Units 1 and 2 B 3.4.14-8 Revision 242 BASES ACTIONS D.1 and D.2 (continued)

(continued) reactor must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits 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, equilibrium xenon, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

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.

F F

TBD

RCS Operational LEAKAGE B 3.4.14 Prairie Island Units 1 and 2 B 3.4.14-9 Revision 242 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by monitoring containment atmosphere radioactivity. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.16, RCS Leakage Detection Instrumentation.

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

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

SR 3.4.14.2 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 LCO 3.4.19, Steam Generator Tube Integrity, should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference

4. 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