ML26034C213
| ML26034C213 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 02/12/2026 |
| From: | Marshall M Plant Licensing Branch 1 |
| To: | Mudrick C Constellation Energy Generation |
| Klett, AL | |
| References | |
| EPID L-2022-LLA-0140 | |
| Download: ML26034C213 (0) | |
Text
February 12, 2026 Mr. Christopher H. Mudrick, Sr.
Senior Vice President and Chief Nuclear Officer Constellation Energy Generation, LLC President Constellation Nuclear 200 Exelon Way Kennett Square, PA 19348
SUBJECT:
LIMERICK GENERATING STATION, UNITS 1 AND 2 - CORRECTION OF AMENDMENT NOS. 268 AND 230 TO REVISE THE LICENSING AND DESIGN BASIS RELATED TO THE REPLACEMENT OF SAFETY-RELATED ANALOG CONTROL SYSTEMS WITH A SINGLE DIGITAL PLANT PROTECTION SYSTEM (EPID L-2022-LLA-0140)
Dear Mr. Mudrick:
On January 2, 2026, the U.S. Nuclear Regulatory Commission (NRC) issued Amendment Nos. 268 and 230 (Agencywide Documents Access and Management System Accession No. ML25325A355) to Renewed Facility Operating License Nos. NPF-39 and NPF-85 for the Limerick Generating Station, Units 1 and 2 (Limerick), respectively. The amendments revised the license and various technical specifications (TSs) in order for the licensee to implement a planned modification that will replace existing safety-related analog control systems with a single digital plant protection system.
The purpose of this letter is to issue corrected license pages, TS pages, and identify additional TS pages that need to be removed from Unit 2 TS. The NRC staff has determined that these are typographical errors and does not change the staffs previous conclusion in the safety evaluation for issuance of Amendment Nos. 268 and 230, nor does it affect the NRC staffs no significant hazards consideration determination.
Unit 1 Corrected License and TS Pages Corrected Page Typographical Error 8
Missing revision bar.
9 Missing revision bar.
1-8 Missing revision bar.
3/4 3-5 In Function 5, ( before superscript n) should be superscript, and superscript (q) should have been deleted.
3/4 3-6 In Function 10, 132.2 should be 132.3.
3/4 3-7 In Function 25, superscript (i) should be (l).
3/4 3-12 In Table Notation (w), one should be one or.
Unit 1 Corrected License and TS Pages Corrected Page Typographical Error 3/4 4-9 Missing revision bar.
3/4 10-8 In LCO 3.10.8, a.3.3.2 should be a.3.3.1.
Unit 2 Corrected License and TS Pages Corrected Page Typographical Error 8
Missing revision bar.
9 Missing revision bar.
1-8 Missing revision bar.
3/4 3-5 In Function 5, ( before superscript n) should be superscript, and superscript (q) should have been deleted.
3/4 3-6 In Function 10, 132.2 should be 132.3.
3/4 3-7 In Function 25, superscript (i) should be (l).
3/4 3-12 In Table Notation (w), one should be one or.
3/4 10-9 In LCO 3.10.8, a.3.3.2 should be a.3.3.1.
In addition to the corrected pages, the following pages need to be removed from Unit 2 TS:
2-4a 3/4 3-1a 3/4 3-8a 3/4 3-36a 3/4 3-41a through 3/4 3-41e Enclosed please find the corrected license and TS pages listed above and the corrected errata sheet for Amendment No. 230.
If you have any questions, please contact me at 301-415-2781 or michael.marshall@nrc.gov.
Sincerely,
/RA/
Michael L. Marshall, Jr., Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.: 50-352 and 50-353
Enclosures:
As stated cc: Listserv
ENCLOSURE LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 CORRECTED LICENSE AND TECHNICAL SPECIFICATION PAGES FOR FACILITY OPERATING LICENSE NOS. NPF-39 and NPF-85 Renewed License No. NPF-39 Amendment No. 255, 268 D.
The facility requires exemptions from certain requirements of 10 CFR Part 50.
These include (a) exemption from the requirement of Appendix J, the testing of containment air locks at times when the containment integrity is not required (Section 6.2.6.1 of the SER and SSER-3), (b) exemption from the requirements of Appendix J, the leak rate testing of the Main Steam Isolation Valves (MSIVs) at the peak calculated containment pressure, Pa, and exemption from the requirements of Appendix J that the measured MSIV leak rates be included in the summation for the local leak rate test (Section 6.2.6 of SSER-3), (c) exemption from the requirement of Appendix J, the local leak rate testing of the Traversing Incore Probe Shear Valves (Section 6.2.6 of the SER and SSER-3).
(25)
The licensees UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as revised in accordance with license condition 2.C.(24),
describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).
(a)
Constellation Energy Generation, LLC shall implement those new programs and enhancements to existing programs no later than April 26, 2024.
(b)
Constellation Energy Generation, LLC shall complete those activities designated for completion prior to the PEO, as noted in Commitment Nos. 18, 19, 20, 22, 23, 24, 28, 29, 30, 38, 39, 40, 41, 42, 43, and 47, of Appendix A of NUREG-2171, Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2, no later than April 26, 2024, or the end of the last refueling outage prior to the period of extended operation, whichever occurs later.
(c)
Constellation Energy Generation, LLC shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be completed in item (b) above.
(26)
Equipment Qualification Testing and Analysis - Plant Protection System Components Prior to startup following the first refueling outage during which the digital instrumentation and control Plant Protection System at the Limerick Generating Station, Unit 1 is installed, Constellation Energy Generation, LLC (CEG) shall complete seismic, environmental, and electromagnetic capability testing and analysis of critical hardware components, as described in CEG letter to the U.S. Nuclear Regulatory Commission, "Response to Requests for Additional Information - Equipment Qualification of Components - Limerick Generating Station Digital Plant Protection System," dated February 21, 2025, and July 10, 2025,, "Response to RAI-37 and -39 through -41" (ADAMS Accession No. ML25191A224). To satisfy this License Condition, CEG will formally document the successful completion of the required equipment qualification testing and analyses that are described in Attachment 1 of the July 10, 2025, submittal.
Renewed License No. NPF-39 Amendment No. 255, 268 E.
Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, submitted by letter dated May 17, 2006, is entitled: Limerick Generating Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2. The set contains Safeguards Information protected under 10 CFR 73.21.
Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The CSP was approved by License Amendment No. 204 and modified by License Amendment No. 218.
F.
Deleted G.
The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
H.
This renewed license is effective as of the date of issuance and shall expire at midnight on October 26, 2044.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
William M. Dean, Director Office of Nuclear Reactor Regulation Attachments/Appendices:
- 1. Attachments 1-2
- 2. Appendix A - Technical Specifications
- 3. Appendix B - Environmental Protection Plan
- 4. Appendix C - Additional Conditions Date of Issuance: October 20, 2014 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.
Therefore these exemptions are hereby granted pursuant to 10 CFR 50.12 and 50.47(c). With the granting of these exemptions the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provision of the Act, and the rules and regulations of the Commission.
DEFINITIONS SHUTDOWN MARGIN (SDM) (Continued) c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-1a.
SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
STAGGERED TEST BASIS 1.42 A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TURBINE BYPASS SYSTEM RESPONSE TIME 1.43A The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required position.
The response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured.
UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
UNRESTRICTED AREA 1.45 UNRESTRICTED AREA means an area, access to which is neither limited nor controlled by the licensee.
LIMERICK - UNIT 1 1-8 Amendment No. 52, 187, 268
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS FUNCTION APPLICABLE OPERATIONAL CONDITIONS MINIMUM OPERABLE CHANNELS ALLOWABLE VALUE ACTION
- 3. Reactor Vessel Pressure (Continued)
- c. Reactor Vessel Pressure -
Low
- 1. LOCA (Permissive) 1,2,3 3
435 psig (decreasing)
- 2. Core Spray (Permissive) 1,2,3 4
435 psig (decreasing) 17
- d. HPCI Steam Supply Pressure
- Low 1,2,3 3
90 psig 4
- e. RCIC Steam Supply Pressure
- Low 1,2,3 3
56.5 psig 4
Wide Range
- a. Low, Low, Low Level 1 1,2(n),3(n) 3
- 136 inches
- b. Low, Low - Level 2 1,2(o)(q),3(o)(q) 3
- 45 inches
- c. High, Level 8 1,2(o)(q),3(o)(q) 4(r) 60 inches 18
Narrow Range a Low - Level 3 1,2(n),3 (n) 3 11.0 inches Reactor Trip System
- 6. Scram Discharge Volume Water Level - High
- a. Level Transmitter 1,2,5(f) 3 261' 5 5/8" elevation
- b. Float Switch 1,2,5(f) 3 261' 5 5/8" elevation
- 7. Reactor Mode Switch Position 1,2, 3
N.A.
3,4, 15 5(w) 16 Drywell
- 8. Drywell Pressure - High 1(p),2(n)(o)(p)(s),
3(n)(o)(p) 3 1.88 psig
- 9. Primary Containment Instrument Gas Line to Drywell Pressure - Low 1,2,3 1/valve 1.9 psi 5
LIMERICK - UNIT 1 3/4 3-5 Amendment No. 268
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS FUNCTION APPLICABLE OPERATIONAL CONDITIONS MINIMUM OPERABLE CHANNELS ALLOWABLE VALUE ACTION Emergency Core Cooling System
- 10. Condensate Storage Tank Level
- Low 1,2(o),3(o)(q) 3 164.3
- inches, 132.3 inches(t) 13
- 11. Automatic Depressurization System (Permissives) 1,2(n),3(n) 6 125 psig, (increasing),
115 psig, (increasing)(u) 18
- 12. LPCI Injection Valve Differential Pressure-Low (Permissive) 1,2,3 1/valve 64 psid and 84 psid 7
High Pressure Coolant Injection (HPCI)
- 13. Suppression Pool Water Level
- High 1,2(o),3(o) 2(r) 24 feet 3 inches 8
- 14. HPCI Steam Line Pressure -
High 1,2,3 2(r) 984" H20 4
20 psig 4
- 16. HPCI Equipment Room Temperature - High 1,2,3 2(r) 177°F, 191°F 4
- 17. HPCI Equipment Room Temperature High 1,2,3 2(r) 108.5°F 4
- 18. HPCI Pipe Routing Area Temperature - High 1,2,3 8
177°F, 191°F 4
Main Steam, Turbine, Condenser
- 19. Main Steam Line Isolation Valve - Closure 1(g) 3 12% closed 14
- 20. Turbine Stop Valve - Closure 1(h) 3 7% closed 1, 11
- 21. Turbine Control Valve Fast Closure, Trip Oil Pressure -
Low 1(h) 3 465 psig 1, 11 LIMERICK - UNIT 1 3/4 3-6 Amendment No. 268
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS FUNCTION APPLICABLE OPERATIONAL CONDITIONS MINIMUM OPERABLE CHANNELS ALLOWABLE VALUE ACTION Main Steam, Turbine, Condenser (Continued)
- 22. Main Steam Line Pressure -
Low 1
3 821 psig 14
- 23. Main Steam Line Flow - High 1,2,3 3/steam line 123 psid
- 24. Condenser Vacuum - Low 1,2(i),3(i) 3 10.1 psia 3
- 25. Outboard MSIV Room Temperature - High 1,2,3 3
10.9 psia 200°F(l) 3 Reactor Water Cleanup System and Standby Liquid Control
- 26. RWCS Flow - High 1,2,3 2(r) 65.2 gpm 4
- 27. RWCS Area Temperature - High 1,2,3 12 160°F or 125°F(j) 4
- 28. RWCS Area Ventilation Temperature - High 1,2,3 12 60°F or 40°F(j) 4
- 29. SLCS Initiation(v) 1,2,3 N.A.
N.A.
4 Reactor Core Isolation Cooling (RCIC)
- 30. RCIC Steam Line Pressure -
High 1,2,3 2(r) 381" H2O 4
20.0 psig 4
- 32. RCIC Equipment Room Temperature - High 1,2,3 2(r) 161°F, 191°F 4
- 33. RCIC Equipment Room Temperature - High 1,2,3 2(r) 113.5°F 4
- 34. RCIC Pipe Routing Area Temperature - High 1,2,3 10 161°F, 191°F 4
LIMERICK - UNIT 1 3/4 3-7 Amendment No. 268
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS TABLE NOTATIONS
Wide range accident monitor per Specification 3.3.7.5.
The Automatic Depressurization System Initiation Function is only required to be OPERABLE when reactor steam dome pressure is 100 psig.
Ž The High Pressure Coolant Injection System initiation functions are only required to be OPERABLE when reactor steam dome pressure is 200 psig.
The High Pressure Coolant Injection System initiation function for Drywell Pressure - High is not required to be OPERABLE when reactor steam dome pressure is < 550 psig.
The Reactor Core Isolation Cooling System initiation functions are only required to be OPERABLE when reactor steam dome pressure is > 150 psig.
A required channel may be placed in bypass for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for surveillance testing provided at least one OPERABLE channel for the same function is monitoring that parameter and is capable of completing its safety function.
This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
The higher Allowable Value is for OPERABILITY of the High Pressure Coolant Injection System. The lower Allowable Value is for OPERABILITY of the
Reactor Core Isolation Cooling System.
The higher Allowable Value is for the OPERABILITY of the Core Spray Pump Discharge Pressure - High Permissive. The lower Allowable Value is for OPERABILITY of the RHR LPCI Mode Pump Discharge Pressure - High Permissive.
For a period of 30 days preceding exit of OPERATIONAL CONDITION 1 at the
start of the 2026 refueling outage, the Reactor Water Cleanup System
Isolation Trip Function is not required to be OPERABLE.
With any control rod withdrawn from a core cell containing one Ž
more fuel
assemblies.
LIMERICK - UNIT 1 3/4 3-12 Amendment No. 268
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
30 gpm total leakage.
d.
25 gpm total leakage averaged over any 24-hour period.
e.
1 gpm leakage at a reactor coolant system pressure of 950 +/-10 psig from any reactor coolant system pressure isolation valve.**
f.
2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, isolate affected component, pipe, or vessel from the reactor coolant system by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With any reactor coolant system leakage greater than the limits in b, c and/or d above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one other closed manual, deactivated automatic, or check* valves, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
With any reactor coolant system leakage greater than the limit in f above, identify the source of leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.
- Pressure isolation valve leakage is not included in any other allowable operational leakage specified in Section 3.4.3.2.
LIMERICK - UNIT 1 3/4 4-9 Amendment No. 28, 49, 172, 182, 254, 268
SPECIAL TEST EXCEPTIONS 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING LIMITING CONDITIONS FOR OPERATION 3.10.8 When conducting inservice leak or hydrostatic testing, the average reactor coolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased to greater than 200°F, and operation considered not to be in OPERATIONAL CONDITION 3:
For performance of an inservice leak or hydrostatic test, As a consequence of maintaining adequate pressure for an inservice leak or hydrostatic test, or As a consequence of maintaining adequate pressure for control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test, provided the following OPERATIONAL CONDITION 3 Specifications are met:
- a. 3.3.1 PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS Functions 4.b, 36, 37, and 38 of Table 3.3.1-1;
- b. 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY;
- c. 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY;
- d. 3.6.5.2.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES;
- e. 3.6.5.2.2 REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES; and
- f. 3.6.5.3 STANDBY GAS TREATMENT SYSTEM.
APPLICABILITY: OPERATIONAL CONDITION 4, with average reactor coolant temperature greater than 200°F.
ACTION:
With the requirements of the above Specifications not satisfied:
- 1. Immediately enter the applicable (OPERATIONAL CONDITION 3) action for the affected Specification; or
- 2. Immediately suspend activities that could increase the average reactor coolant temperature or pressure and reduce the average reactor coolant temperature to 200°F or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.10.8 Verify applicable OPERATIONAL CONDITION 3 surveillances for the Specifications listed in 3.10.8 are met.
LIMERICK - UNIT 1 3/4 10-8 Amendment No. 133, 249, 268
The licensees UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as revised in accordance with license condition 2.C.(12),
describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).
D
Constellation Energy Generation, LLC shall implement those new programs and enhancements to existing programs no later than December 22, 2028.
E
Constellation Energy Generation, LLC shall complete those activities designated for completion prior to the PEO, as noted in Commitment Nos. 18, 19, 20, 22, 23, 24, 28, 29, 30, 38, 39, 40, 41, 42, 43, and 47, of Appendix A of NUREG-2171, Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2, no later than December 22, 2028, or the end of the last refueling outage prior to the period of extended operation, whichever occurs later.
F
Constellation Energy Generation, LLC shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be completed in item (b) above.
The Additional Conditions contained in Appendix C, as revised through Amendment No. , are hereby incorporated into this renewed license.
Constellation Energy Generation, LLC shall operate the facility in accordance with the Additional Conditions.
Renewed License No. NPF-85 Amendment No. 217, 217 230 (15)
Equipment Qualification Testing and Analysis - Plant Protection System Components Prior to startup following the first refueling outage during which the digital instrumentation and control Plant Protection System at the Limerick Generating Station, Unit 2 is installed, Constellation Energy Generation, LLC shall complete seismic, environmental, and electromagnetic capability testing and analysis of critical hardware components, as described in CEG letter to the U.S. Nuclear Regulatory Commission, "Response to Requests for Additional Information - Equipment Qualification of Components - Limerick Generating Station Digital Plant Protection System" dated February 21, 2025 and July 10, 2025,, "Response to RAI-37 and -39 through -41" (ADAMS Accession No. ML25191A224). To satisfy this License Condition, CEG will formally document the successful completion of the required equipment qualification testing and analyses that are described in of the July 10, 2025 submittal.
E.
Deleted F.
The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
G.
This renewed license is effective as of the date of issuance and shall expire at midnight on June 22, 2049.
Enclosures:
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
William M. Dean, Director Office of Nuclear Reactor Regulation
- 1. Appendix A -Technical Specifications
- 2. Appendix B - Environmental Protection Plan
- 3. Appendix C -Additional Conditions Date of Issuance: October 20, 2014 Renewed License No. NPF-85 Amendment No. 193, 230 D.
The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include (a) exemption from the requirement of Appendix J, the testing of containment air locks at times when the containment integrity is not required (Section 6.2.6.1 of the SER and SSER-3), (b) exemption from the requirements of Appendix J, the leak rate testing of the Main Steam Isolation Valves (MSIVs) at the peak calculated containment pressure, Pa, and exemption from the requirements of Appendix J that the measured MSIV leak rates be included in the summation for the local leak rate test (Section 6.2.6.1 of SSER-3), (c) exemption from the requirement of Appendix J, the local leak rate testing of the Traversing Incore Probe Shear Valves (Section 6.2.6.1 of the SER and SSER-3), and (d) an exemption from the schedule requirements of 10 CFR 50.33(k)(l) related to availability of funds for decommissioning the facility (Section 22.1, SSER 8). The special circumstances regarding exemptions (a),
(b) and (c) are identified in Sections 6.2.6.1 of the SER and SSER 3. An exemption from the criticality monitoring requirements of 10 CFR 70.24 was previously granted with NRC materials license No. SNM-1977 issued November 22, 1988. The licensee is hereby exempted from the requirements of 10 CFR 70.24 insofar as this requirement applies to the handling and storage of fuel assemblies held under this renewed license.
DEFINITIONS SHUTDOWN MARGIN (SDM) (Continued) c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-1a.
SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
STAGGERED TEST BASIS 1.42 A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TURBINE BYPASS SYSTEM RESPONSE TIME 1.43A The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required position.
The response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured.
UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
UNRESTRICTED AREA 1.45 UNRESTRICTED AREA means an area, access to which is neither limited nor controlled by the licensee.
LIMERICK - UNIT 2 1-8 Amendment No. 16, 148, 230
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS FUNCTION APPLICABLE OPERATIONAL CONDITIONS MINIMUM OPERABLE CHANNELS ALLOWABLE VALUE ACTION
- 3. Reactor Vessel Pressure (Continued)
- c. Reactor Vessel Pressure -
Low
- 1. LOCA (Permissive) 1,2,3 3
435 psig (decreasing)
- 2. Core Spray (Permissive) 1,2,3 4
435 psig (decreasing) 17
- d. HPCI Steam Supply Pressure
- Low 1,2,3 3
90 psig 4
- e. RCIC Steam Supply Pressure
- Low 1,2,3 3
56.5 psig 4
Wide Range
- a. Low, Low, Low Level 1 1,2(n),3(n) 3
- 136 inches
- b. Low, Low - Level 2 1,2(o)(q),3(o)(q) 3
- 45 inches
- c. High, Level 8 1,2(o)(q),3(o)(q) 4(r) 60 inches 18
Narrow Range a Low - Level 3 1,2(n),3(n) 3 11.0 inches Reactor Trip System
- 6. Scram Discharge Volume Water Level - High
- a. Level Transmitter 1,2,5(f) 3 261' 5 5/8" elevation
- b. Float Switch 1,2,5(f) 3 261' 5 5/8" elevation
- 7. Reactor Mode Switch Position 1,2, 3
N.A.
3,4, 15 5(w) 16 Drywell
- 8. Drywell Pressure - High 1(p),2(n)(o)(p)(s),
3(n)(o)(p) 3 1.88 psig
- 9. Primary Containment Instrument Gas Line to Drywell Pressure - Low 1,2,3 1/valve 1.9 psi 5
LIMERICK - UNIT 2 3/4 3-5 Amendment No. 230
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS FUNCTION APPLICABLE OPERATIONAL CONDITIONS MINIMUM OPERABLE CHANNELS ALLOWABLE VALUE ACTION Emergency Core Cooling System
- 10. Condensate Storage Tank Level
- Low 1,2(o),3(o)(q) 3 164.3
- inches, 132.3 inches(t) 13
- 11. Automatic Depressurization System (Permissives) 1,2(n),3(n) 6 125 psig, (increasing),
115 psig, (increasing)(u) 18
- 12. LPCI Injection Valve Differential Pressure-Low (Permissive) 1,2,3 1/valve 64 psid and 84 psid 7
High Pressure Coolant Injection (HPCI)
- 13. Suppression Pool Water Level
- High 1,2(o),3(o) 2(r) 24 feet 3 inches 8
- 14. HPCI Steam Line Pressure -
High 1,2,3 2(r) 984" H20 4
20 psig 4
- 16. HPCI Equipment Room Temperature - High 1,2,3 2(r) 177°F, 191°F 4
- 17. HPCI Equipment Room Temperature High 1,2,3 2(r) 108.5°F 4
- 18. HPCI Pipe Routing Area Temperature - High 1,2,3 8
177°F, 191°F 4
Main Steam, Turbine, Condenser
- 19. Main Steam Line Isolation Valve - Closure 1(g) 3 12% closed 14
- 20. Turbine Stop Valve - Closure 1(h) 3 7% closed 1, 11
- 21. Turbine Control Valve Fast Closure, Trip Oil Pressure -
Low 1(h) 3 465 psig 1, 11 LIMERICK - UNIT 2 3/4 3-6 Amendment No. 230
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS FUNCTION APPLICABLE OPERATIONAL CONDITIONS MINIMUM OPERABLE CHANNELS ALLOWABLE VALUE ACTION Main Steam, Turbine, Condenser (Continued)
- 22. Main Steam Line Pressure -
Low 1
3 821 psig 14
- 23. Main Steam Line Flow - High 1,2,3 3/steam line 123 psid
- 24. Condenser Vacuum - Low 1,2(i),3(i) 3 10.1 psia 3
- 25. Outboard MSIV Room Temperature - High 1,2,3 3
10.9 psia 200°F(l) 3 Reactor Water Cleanup System and Standby Liquid Control
- 26. RWCS Flow - High 1,2,3 2(r) 65.2 gpm 4
- 27. RWCS Area Temperature - High 1,2,3 12 160°F or 125°F(j) 4
- 28. RWCS Area Ventilation Temperature - High 1,2,3 12 60°F or 40°F(j) 4
- 29. SLCS Initiation(v) 1,2,3 N.A.
N.A.
4 Reactor Core Isolation Cooling (RCIC)
- 30. RCIC Steam Line Pressure -
High 1,2,3 2(r) 381" H2O 4
20.0 psig 4
- 32. RCIC Equipment Room Temperature - High 1,2,3 2(r) 161°F, 191°F 4
- 33. RCIC Equipment Room Temperature - High 1,2,3 2(r) 113.5°F 4
- 34. RCIC Pipe Routing Area Temperature - High 1,2,3 10 161°F, 191°F 4
LIMERICK - UNIT 2 3/4 3-7 Amendment No. 230
TABLE 3.3.1-1 (Continued)
PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS TABLE NOTATIONS (m)
Wide range accident monitor per Specification 3.3.7.5.
(n)
The Automatic Depressurization System Initiation Function is only required to be OPERABLE when reactor steam dome pressure is 100 psig.
(o)
The High Pressure Coolant Injection System initiation functions are only required to be OPERABLE when reactor steam dome pressure is 200 psig.
(p)
The High Pressure Coolant Injection System initiation function for Drywell Pressure - High is not required to be OPERABLE when reactor steam dome pressure is < 550 psig.
(q)
The Reactor Core Isolation Cooling System initiation functions are only required to be OPERABLE when reactor steam dome pressure is > 150 psig.
(r)
A required channel may be placed in bypass for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for surveillance testing provided at least one OPERABLE channel for the same function is monitoring that parameter and is capable of completing its safety function.
(s)
This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
(t)
The higher Allowable Value is for OPERABILITY of the High Pressure Coolant Injection System. The lower Allowable Value is for OPERABILITY of the Reactor Core Isolation Cooling System.
(u)
The higher Allowable Value is for the OPERABILITY of the Core Spray Pump Discharge Pressure - High Permissive. The lower Allowable Value is for OPERABILITY of the RHR LPCI Mode Pump Discharge Pressure - High Permissive.
(v)
For a period of 30 days preceding exit of OPERATIONAL CONDITION 1 at the start of the 2027 refueling outage, the Reactor Water Cleanup System Isolation Trip Function is not required to be OPERABLE.
(w)
With any control rod withdrawn from a core cell containing one or more fuel assemblies.
LIMERICK - UNIT 2 3/4 3-12 Amendment No. 230
SPECIAL TEST EXCEPTIONS 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING LIMITING CONDITION FOR OPERATION 3.10.8 When conducting inservice leak or hydrostatic testing, the average reactor coolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased to greater than 200°F, and operation considered not to be in OPERATIONAL CONDITION 3:
For performance of an inservice leak or hydrostatic test, As a consequence of maintaining adequate pressure for an inservice leak or hydrostatic test, or As a consequence of maintaining adequate pressure for control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test, provided the following OPERATIONAL CONDITION 3 Specifications are met:
- a. 3.3.1 PLANT PROTECTION SYSTEM INSTRUMENTATION CHANNELS Functions 4.b, 36, 37, and 38 of Table 3.3.1-1;
- b. 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY;
- c. 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY;
- d. 3.6.5.2.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES;
- e. 3.6.5.2.2 REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES; and
- f. 3.6.5.3 STANDBY GAS TREATMENT SYSTEM.
APPLICABILITY: OPERATIONAL CONDITION 4, with average reactor coolant temperature greater than 200°F.
ACTION:
With the requirements of the above Specifications not satisfied:
- 1. Immediately enter the applicable (OPERATIONAL CONDITION 3) action for the affected Specification; or
- 2. Immediately suspend activities that could increase the average reactor coolant temperature or pressure and reduce the average reactor coolant temperature to 200°F or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.10.8 Verify applicable OPERATIONAL CONDITION 3 surveillances for the Specifications listed in 3.10.8 are met.
LIMERICK - UNIT 2 3/4 10-9 Amendment No. 95, 211, 230
ENCLOSURE LIMERICK GENERATING STATION, UNIT 2 DOCKET NO. 50-353 CORRECTED ERRATA SHEET FOR AMENDMENT NO. 230 FOR FACILITY OPERATING LICENSE NO. NPF-85
ATTACHMENT TO LICENSE AMENDMENT NO. 230 LIMERICK GENERATING STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following pages of Renewed Facility Operating License with the revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Page Insert Page 3
3 8
8 9
9 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
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- via eConcurrence NRR-106 OFFICE NRR/DORL/LPL1/PM*
NRR/DORL/LPL2-2/LA*
NRR/DEX/EICB/BC*
NAME MMarshall ABaxter FSacko DATE 2/4/2026 2/9/2026 2/9/2026 OFFICE NRR/DORL/LPL2-1/BC*
NRR/DORL/LPL1/PM*
NAME UShoop MMarshall DATE 2/11/2026 2/12/2026