ML23010A108
| ML23010A108 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/15/2023 |
| From: | Audrey Klett Plant Licensing Branch 1 |
| To: | David Jones Susquehanna |
| Klett A | |
| References | |
| EPID L-2023-LLA-0003 | |
| Download: ML23010A108 (1) | |
Text
January 15, 2023 Mr. Derek Jones Acting Site Vice President Susquehanna Nuclear, LLC 769 Salem Boulevard NUCSB3 Berwick, PA 18603-0467
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 - ISSUANCE OF AMENDMENT NO. 268 RE: CHANGE TO CERTAIN TECHNICAL SPECIFICATIONS FOR CONTROL RODS (EMERGENCY CIRCUMSTANCES)
Dear Mr. Jones:
The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 268 to Renewed Facility Operating License No. NPF-22 for the Susquehanna Steam Electric Station (Susquehanna), Unit 2. The amendment consists of changes to the technical specifications (TS) in response to Susquehanna Nuclear, LLCs application dated January 10, 2023, as supplemented by letter dated January 14, 2023.
The amendment revises Susquehanna Unit 2s TS 3.1.3, Control Rod OPERABILITY, TS 3.1.6, Rod Pattern Control, and TS 3.3.2.1, Control Rod Block Instrumentation, by adding references to the analyzed rod position sequence to temporarily allow for greater flexibility in rod manipulation during various stages of reactor power operation.
The license amendment is issued under emergency circumstances as provided in the provisions of paragraph 50.91(a)(5) of Title 10 of the Code of Federal Regulations because of the time-critical nature of the amendment. In this instance, an emergency situation exists in that the amendment is needed to allow the licensee to startup following a maintenance outage.
The NRCs related safety evaluation is enclosed. The safety evaluation describes the emergency circumstances under which the amendment is issued and the final no significant hazards determination. The NRC will include a notice of issuance addressing the final no
significant hazards determination and opportunity for a hearing associated with the emergency circumstances in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Audrey Klett, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-388
Enclosures:
- 1. Amendment No. 268 to NPF-22
- 2. Safety Evaluation cc: Listserv SUSQUEHANNA NUCLEAR, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.
DOCKET NO. 50-388 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. NPF-22
- 1.
The U.S. Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for the amendment filed by Susquehanna Nuclear, LLC, dated January 10, 2023, as supplemented by letter dated January 14, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations, and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-22 is hereby amended to read as follows:
2.C.(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 268, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Susquehanna Nuclear, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented as soon as practicable.
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 15, 2023 Hipolito J.
Gonzalez Digitally signed by Hipolito J. Gonzalez Date: 2023.01.15 13:44:33 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 268 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following page of Renewed Facility Operating License No. NPF-22 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT Page 3 Page 3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
REMOVE INSERT 3.1-8 3.1-8 3.1-9 3.1-9 3.1-18 3.1-18 3.1-19 3.1-19 3.3-17 3.3-17 3.3-18 3.3-18 3.3-19 3.3-19 3.3-20 3.3-20
Renewed Operating License No. NPF-22 Amendment No. 268 (3)
Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, posses, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, posses, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
Susquehanna Nuclear, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission nor or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Susquehanna Nuclear, LLC is authorized to operate the facility at reactor core power levels not in excess of 3952 megawatts thermal in accordance with the conditions specified herein. The preoperational tests, startup tests and other items identified in License Conditions 2.C.(20), 2.C.(21), 2.C.(22), and 2.C.(23) to this license shall be completed as specified.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 268, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Susquehanna Nuclear, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
For Surveillance Requirements (SRs) that are new in Amendment 151 to Facility Operating License No. NPF-22, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 151. For SRs that existed prior to Amendment 151, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 151.
Control Rod OPERABILITY 3.1.3 SUSQUEHANNA - UNIT 2 3.1-8 Amendment 151, 229, 268 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.3 Perform SR 3.1.3.3 for each withdrawn OPERABLE control rod.
AND A.4 Perform SR 3.1.1.1.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B.
Two or more withdrawn control rods stuck.
B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C.
One or more control rods inoperable for reasons other than Condition A or B.
C.1 ----------------NOTE----------------
RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.
Fully insert inoperable control rod.
AND C.2 Disarm the associated CRD.
3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 4 hours
Control Rod OPERABILITY 3.1.3 SUSQUEHANNA - UNIT 2 3.1-9 Amendment 151, 268 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D.
NOTE------------
Not applicable when THERMAL POWER
> 10% RTP.
Two or more inoperable control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.
D.1 Restore compliance with BPWS.
OR D.2 Restore control rod to OPERABLE status.
OR D.3 Confirm compliance with the analyzed rod position sequence.1 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> E.
NOTE------------
Not applicable when THERMAL POWER
> 10% RTP.
One or more BPWS groups with four or more inoperable control rods.
E.1 Restore control rod to OPERABLE status.
OR E.2 Confirm compliance with the analyzed rod position sequence.1 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours F.
Required Action and associated Completion Time of Condition A, C, D, or E not met.
OR Nine or more control rods inoperable.
F.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1 This Required Action is only applicable during the remainder of Unit 2, Cycle 21. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
Rod Pattern Control 3.1.6 SUSQUEHANNA - UNIT 2 3.1-18 Amendment 151, 268 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS).
NOTE--------------------------------------------------------------
For Unit 2, Cycle 21 only, OPERABLE control rods may comply with the requirements of the analyzed rod position sequence in lieu of the banked position withdrawal sequence.1 APPLICABILITY:
MODES 1 and 2 with THERMAL POWER 10% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more OPERABLE control rod(s) not in compliance with BPWS.2 A.1
NOTE-----------------
Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation."
Move associated control rod(s) to correct position.
OR A.2 Declare associated control rod(s) inoperable.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 hours 1 This Note is only applicable during the remainder of Unit 2, Cycle 21. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
2 During Unit 2, Cycle 21 only, one or more OPERABLE control rods not in compliance with the analyzed rod position sequence requires entry into Condition A. The Required Actions remain unchanged except that Required Action A.1 refers to the correct position per the analyzed rod position sequence in lieu of BPWS. Upon completion of Unit 2, Cycle 21, this temporary requirement is no longer applicable and will expire on April 15, 2023.
Rod Pattern Control 3.1.6 SUSQUEHANNA - UNIT 2 3.1-19 Amendment 151, 247, 268 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B.
Nine or more OPERABLE control rods not in compliance with BPWS. 1 B.1
NOTE-------------------
Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1.
Suspend withdrawal of control rods.
AND B.2 Place the reactor mode switch in the shutdown position.
Immediately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 During Unit 2, Cycle 21 only, nine or more OPERABLE control rods not in compliance with the analyzed rod position sequence requires entry into Condition B. The Required Actions remain unchanged. Upon completion of Unit 2, Cycle 21, this temporary requirement is no longer applicable and will expire on April 15, 2023.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with BPWS. 2 In accordance with the Surveillance Frequency Control Program 2 During Unit 2, Cycle 21 only, verification of compliance with the analyzed rod position sequence may be performed in lieu of compliance with the BPWS to meet SR 3.1.6.1. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
Control Rod Block Instrumentation 3.3.2.1 SUSQUEHANNA - UNIT 2 3.3-17 Amendment 151, 265, 268 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C.
Rod worth minimizer (RWM) inoperable during reactor startup.
C.1 Suspend control rod movement except by scram.
OR C.2.1.1 Verify 12 rods withdrawn.
OR C.2.1.2 Verify by administrative methods that startup with RWM inoperable has not been performed in the last calendar year.
AND C.2.2 Verify movement of control rods is in compliance with banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.1 Immediately Immediately Immediately During control rod movement D.
RWM inoperable during reactor shutdown.
D.1 Verify movement of control rods is in accordance with BPWS by a second licensed operator or other qualified member of the technical staff.1 During control rod movement 1 During Unit 2, Cycle 21 only, verification of compliance with the analyzed rod position sequence may be performed in lieu of verification of compliance with BPWS to meet Required Actions C.2.2 and D.1. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
Control Rod Block Instrumentation 3.3.2.1 SUSQUEHANNA - UNIT 2 3.3-18 Amendment 151, 207, 220, 247, 265, 268 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E.
One or more Reactor Mode Switch-Shutdown Position channels inoperable.
E.1 Suspend control rod withdrawal.
AND E.2 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.
Immediately Immediately SURVEILLANCE REQUIREMENTS
NOTES----------------------------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
- 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains control rod block capability.
SURVEILLANCE FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.2
NOTE--------------------------------
Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn at 10% RTP in MODE 2.
Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program
Control Rod Block Instrumentation 3.3.2.1 SUSQUEHANNA - UNIT 2 3.3-19 Amendment 151, 220, 247, 265, 268 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.2.1.3
NOTE--------------------------------
Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is 10% RTP in MODE 1.
Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.4 Verify the RBM:
- a. Low Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is 28% RTP and Intermediate Power Range Setpoint specified in the COLR.
- b. Intermediate Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is > Intermediate Power Range Setpoint specified in the COLR and High Power Range Setpoint specified in the COLR.
- c. High Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power
> High Power Range Setpoint specified in the COLR.
In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.5 Verify the RWM is not bypassed when THERMAL POWER is 10% RTP.
In accordance with the Surveillance Frequency Control Program
Control Rod Block Instrumentation 3.3.2.1 SUSQUEHANNA - UNIT 2 3.3-20 Amendment 151, 207, 220, 265, 268 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.2.1.6
NOTE--------------------------------
Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after reactor mode switch is in the shutdown position.
Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.7
NOTE--------------------------------
Neutron detectors are excluded.
Perform CHANNEL CALIBRATION.
In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.8 Verify control rod sequences input to the RWM are in conformance with BPWS.1 Prior to declaring RWM OPERABLE following loading of sequence into RWM 1 During Unit 2, Cycle 21 only, verification of compliance with the analyzed rod position sequence may be performed in lieu of verification of compliance with BPWS to meet SR 3.3.2.1.8. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
January 15, 2023 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-22 SUSQUEHANNA NUCLEAR, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.
SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 DOCKET NO. 50-388 Table of Contents
1.0 INTRODUCTION
................................................................................................................ 1 1.1 Background.................................................................................................................... 1 1.2 Description of Control Rod Design Basis and Technical Specifications......................... 1 1.3 Description of the Proposed Changes............................................................................ 3
2.0 REGULATORY EVALUATION
........................................................................................... 4 2.1 Regulatory Requirements............................................................................................... 4 2.2 Licensing Basis............................................................................................................... 5
3.0 TECHNICAL EVALUATION
............................................................................................... 6 3.1 Evaluation of Addition of Analyzed Rod Position Sequence.......................................... 6 3.2 Evaluation of Proposed Changes to TS 3.1.3, 3.1.6, and 3.3.2.1.................................. 7 3.3 Technical Evaluation Conclusion.................................................................................... 7 4.0 EMERGENCY SITUATION................................................................................................ 8
5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
.................................................. 8
6.0 STATE CONSULTATION
................................................................................................ 10
7.0 ENVIRONMENTAL CONSIDERATION
........................................................................... 10
8.0 CONCLUSION
................................................................................................................. 11
9.0 REFERENCES
................................................................................................................. 11 10.0 ABBREVIATIONS............................................................................................................ 12 11.0 PRINCIPAL CONTRIBUTORS........................................................................................ 12
1.0 INTRODUCTION
1.1 Background
By application dated January 10, 2023 [1], as supplemented by letter dated January 14, 2023 [2], Susquehanna Nuclear, LLC (the licensee) submitted a license amendment request (LAR) pertaining to the Susquehanna Steam Electric Station (Susquehanna), Unit 2, technical specifications, which are in Appendix A of the Susquehanna Renewed Facility Operating License No. NPF-22. The proposed changes would revise Susquehanna, Unit 2s Technical Specification (TS) 3.1.3, Control Rod OPERABILITY, TS 3.1.6, Rod Pattern Control, and TS 3.3.2.1, Control Rod Block Instrumentation, by adding references to the analyzed rod position sequence to temporarily allow for greater flexibility in rod manipulation during various stages of reactor power operation. In its application, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC or the Commission) process the proposed amendment under emergency circumstances.
The NRC staff audited various licensee documents and interviewed licensee staff to support the licensing review. The NRC staff issued its audit plan on January 11, 2023 [3], and conducted the audit using an Internet-based portal provided by the licensee and virtual meetings held on January 12 and 13, 2023. The staff issued its audit report on January 15, 2023 [4].
By email dated January 13, 2023 [5], the NRC staff requested additional information from the licensee. The licensee responded to the NRC staffs request by letter dated January 14, 2023 [2].
1.2 Description of Control Rod Design Basis and Technical Specifications Section 4.1.2, Reactor Internal Components, of the updated final safety analysis report (UFSAR) [6] states, in part, that the core (fuel, channels, control rods, and instrumentation) and the core support structure (including the shroud, top guide and core plate) are some of the major reactor internal components. Section 4.1, Summary Description, of the UFSAR [6] states that the reactor assembly includes the control rods, control rod drive housings, and the control rod drives. Each reactor contains 764 fuel assemblies and 185 control rods. Section 4.1.3, Reactivity Control Systems, of the UFSAR [6] states that the control rods perform dual functions of power distribution shaping and reactivity control.
In its LAR [1], the licensee stated that the control rods are components of the control rod drive system, which is the primary reactivity control system for the reactor. The licensee also indicated that the control rod drive system with the reactor protection system provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occurrences, that specified acceptable fuel design limits are not exceeded. The licensee stated that the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the control rod drive system.
Section 15.4.1.2.2.1, Sequence of Events, of the UFSAR [6] states that the purpose of the rod worth minimizer (RWM) is to control rod patterns during startup, such that only specified rod sequences and relative positions are allowed over the operating range from all control rods inserted to approximately 10% of rated core power. The sequences effectively limit the potential amount and rate of reactivity increase during a control rod drop accident (CRDA).
2 In its LAR [1], the licensee indicates that the design basis accident that results in a positive reactivity insertion in a boiling water reactor is the CRDA, which assumes a control rod becomes uncoupled from its control rod drive mechanism prior to or during its withdrawal. Section 15.4.9, Control Rod Drop Accident (CRDA), of the UFSAR [6] states the control rod drop accident is the result of a postulated event in which a high worth control rod is inserted in-sequence into the core. Subsequently, it becomes decoupled from its drive mechanism. The mechanism is withdrawn, but the decoupled control rod is assumed to be stuck in place. At a later optimum moment, the control rod suddenly falls free and drops out of the core. This results in the insertion of large positive reactivity to the core and causes a localized power excursion.
Section 15.4.9.1.2 of the UFSAR [6] states that the CRDA is categorized as a limiting fault because it is not expected to occur during the lifetime of the plant; but, if postulated to occur, it has consequences that include the potential for the release of radioactive material from the fuel.
General design criterion (GDC) 28 in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, requires, in part, reactivity control systems be designed to provide limitations on potential amounts and rates of reactivity increases to ensure that the effects of postulated reactivity accidents do not damage the reactor coolant system, the core, or reactor pressure vessel internals, to ensure long term core cooling capability.
As described in the General Electric Company Report NEDO-21231, Banked Position Withdrawal Sequence [7], the banked position withdrawal sequence (BPWS) mitigates the consequences of a CRDA in the startup and low power operating ranges of a boiling water reactor by reducing control rod reactivity worth. In its LAR [1], the licensee indicates that the BPWS limits the potential reactivity increase from a postulated CRDA during reactor startups and shutdowns below the low power setpoint (LPSP) of 10 % rated thermal power. CRDA analyses assume that the plant operator follows prescribed withdrawal sequences that define the bounding assumptions for the CRDA analysis. The BPWS is applied to both reactor startup and shutdown processes to limit the impact of a CRDA. By using the RWM, which is a rod pattern control system that validates that the BPWS is maintained, the maximum control rod worth during each rod step of the startup or shutdown is maintained within the bounding acceptable levels of the CRDA analysis. By using the RWM and operator actions, as controlled by plant procedures, the proper withdrawal or insertion rod sequence is followed for startup or shutdown evolutions, respectively, and the potential reactivity addition from a CRDA during that evolution is minimized.
Technical specification limiting condition for operation (LCO) 3.1.3 states that each control rod shall be OPERABLE in MODES 1 (power operation) and 2 (startup). If two or more inoperable control rods are not in compliance with the BPWS and not separated by two or more OPERABLE control rods, then Required Actions D.1 and D.2 require the licensee to either restore compliance with BPWS within four hours or restore the control rod to OPERABLE status within four hours. If there are one or more BPWS groups with four or more inoperable control rods, then Required Action E.1 requires the licensee to restore the control rod to OPERABLE status within four hours.
Technical specification LCO 3.1.6 states that OPERABLE control rods shall comply with the requirements of the BPWS in MODES 1 and 2 with thermal power less than or equal to 10 percent rated thermal power. If one or more OPERABLE control rods are not in compliance with BPWS, then Required Actions A.1 and A.2 require the licensee to move the associated control rods to the correction position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated control rod(s) inoperable within eight hours, respectively. If nine or more OPERABLE control rods are not in
3 compliance with BPWS, then Required Actions B.1 and B.2 require the licensee to suspend withdrawal of control rods immediately and place the reactor mode switch in the shutdown position within one hour. Surveillance Requirement 3.1.6.1 requires the licensee to verify all OPERABLE control rods comply with the BPWS in accordance with the surveillance frequency control program.
Technical specification LCO 3.3.2.1 states that the control rod block instrumentation for each function in Table 3.3.2.1-1 shall be OPERABLE according to Table 3.3.2.1-1. If the RWM is inoperable during reactor startup, then Required Actions C.1, C.2.1.1, C.2.1.2, and C.2.2, require, in part, the licensee to suspend control rod movement (except by scram) immediately or: immediately verify at least 12 rods are withdrawn or that startup with the RWM inoperable has not been performed in the last calendar year and that movement of control rods is in compliance with BPWS during control rod movement. If the RWM is inoperable during reactor shutdown, then Required Action D.1 requires, in part, the licensee to verify movement of control rods is in accordance with BPWS curing control rod movement. Surveillance requirement 3.3.2.1.8 requires the licensee to verify control rod sequences input to the RWM are in conformance with BPWS prior to declaring the RWM OPERABLE following loading of sequence into the RWM.
The technical specifications bases pages provided in the LAR [1] state that out-of-sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA.
At less than or equal to 10 percent rated thermal power, the generic BPWS analysis requires inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal. Therefore, if two or more inoperable control rods are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the control rods to OPERABLE status. In addition to the separation requirements for inoperable control rods, a BPWS assumption requires that no more than three inoperable control rods are allowed in any one BPWS group.
1.3 Description of the Proposed Changes In its LAR [1], as supplemented [2], the licensee proposed to modify TS 3.1.3, 3.1.6, and 3.3.2.1 to add references to the analyzed rod position sequence to allow for greater flexibility in rod manipulation during various stages of reactor power operation. The changes would allow the use of alternate requirements on control rod withdrawal order and conditions to protect against a postulated CRDA during startup and low power conditions. Specifically, the licensee requested the following changes:
In TS 3.1.3, add new Required Actions D.3 and E.2 that would be preceded by OR and would state, Confirm compliance with the analyzed rod position sequence.1 The new required actions would both have a completion time of four hours. Footnote 1 would state, This Required Action is only applicable during the remainder of Unit 2, Cycle 21.
Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
In TS 3.1.6, add a note directly under LCO 3.1.6 that would state, For Unit 2, Cycle 21 only, OPERABLE control rods may comply with the requirements of the analyzed rod position sequence in lieu of the banked position withdrawal sequence.1 Footnote 1 would state, This Note is only applicable during the remainder of Unit 2, Cycle 21. Upon
4 completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
In TS 3.1.6, add a new Footnote 2 after Condition A that would state, During Unit 2, Cycle 21 only, one or more OPERABLE control rods not in compliance with the analyzed rod position sequence requires entry into Condition A. The Required Actions remain unchanged except that Required Action A.1 refers to the correct position per the analyzed rod position sequence in lieu of BPWS. Upon completion of Unit 2, Cycle 21, this temporary requirement is no longer applicable and will expire on April 15, 2023.
In TS 3.1.6, add a new Footnote 1 after Condition B that would state, During Unit 2, Cycle 21 only, nine or more OPERABLE control rods not in compliance with the analyzed rod position sequence requires entry into Condition B. The Required Actions remain unchanged. Upon completion of Unit 2, Cycle 21, this temporary requirement is no longer applicable and will expire on April 15, 2023.
In TS 3.1.6, add a new Footnote 2 to Surveillance Requirement 3.1.6.1 that would state, During Unit 2, Cycle 21 only, verification of compliance with the analyzed rod position sequence may be performed in lieu of compliance with the BPWS to meet SR [Surveillance Requirement] 3.1.6.1. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
In TS 3.3.2.1, add a new Footnote 1 to Required Actions C.2.2 and D.1 that would state, During Unit 2, Cycle 21 only, verification of compliance with the analyzed rod position sequence may be performed in lieu of verification of compliance with BPWS to meet Required Actions C.2.2 and D.1. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
In TS 3.3.2.1, add a new Footnote 1 to SR 3.3.2.1.8 that would state, During Unit 2, Cycle 21 only, verification of compliance with the analyzed rod position sequence may be performed in lieu of verification of compliance with BPWS to meet SR 3.3.2.1.8. Upon completion of Unit 2, Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
The licensee proposed this amendment to allow startup following a maintenance outage. The licensee identified a challenge to comply with the BPWS, which would prevent startup following a maintenance outage without significantly expanding the scope of the outage. In its LAR, the licensee indicates that it has been monitoring control rod friction during the current Unit 2 operating cycle (Cycle 21). The licensee determined that the friction is a result of fuel channel deformation on high exposure ATRIUM 10 fuel assemblies, which are primarily located on the periphery of the core where multiple high exposure assemblies are loaded into the same control cell. The licensee indicated that in the startup range below the LPSP, control rod friction may complicate rod withdrawal and, thus, require some rods to remain fully inserted out of sequence.
This presents a challenge to comply with the BPWS requirements. The licensee indicated that in prior startups, it has been able to manage rods that must remain fully inserted during the startup range by reordering the pull sequence, in accordance with BPWS rules, to mitigate the impacts of these out-of-sequence fully inserted rods.
2.0 REGULATORY EVALUATION
2.1 Regulatory Requirements Under Section 50.92(a) of Title 10 to the Code of Federal Regulations (10 CFR), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and
5 appropriate. The common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations. Accordingly, for this LAR, the NRC staff must conclude that there is reasonable assurance that the actions taken when an LCO is not met and the changes to the surveillance requirements do not endanger public health and safety.
Section 50.36, Technical specifications, of 10 CFR establishes the requirements related to the content of the technical specifications. Pursuant to 10 CFR 50.36(c), technical specifications are required, in part, to include LCOs. Section 50.36(c)(2)(i) states, in part, that LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility, and when LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the LCO can be met.
Section 50.36(c)(3), Surveillance requirements, of 10 CFR states:
Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
Appendix A to 10 CFR Part 50 provides the minimum necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. GDC 28, Reactivity Limits, states:
The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
2.2 Licensing Basis The NRC staff considered the following licensing basis during its review:
Amendment No. 260 [8] to Renewed Facility Operating License Nos. NPF-22, which authorized the licensee to apply Framatome analysis methodologies necessary to support a planned transition to ATRIUM 11 fuel. Specifically, the staff considered the amendments approval of the use of the following analytical method in the core operating limits report under TS 5.6.5.b. 22, ANP-10333P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA) [9].
Chapter 15, Accident Analyses, of the UFSAR [6], which examines the effects and consequences of anticipated process disturbances and postulated component failures and evaluates the capability of the units to control or accommodate such failures and
6 events. Specifically, the staff considered section 15.4.9, Control Rod Drop Accident (CRDA).
3.0 TECHNICAL EVALUATION
Under 10 CFR 50.92(a), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The staff evaluated the request to determine whether the proposed changes are consistent with the regulations and licensing basis discussed in section 2.0 of this safety evaluation. The staff reviewed the proposed changes to the technical specifications to determine whether they meet the requirements of 10 CFR 50.36 and GDC 28 and provide reasonable assurance that operation with the new allowable control rod insertion and withdrawal sequences will not endanger the health and safety of the public.
3.1 Evaluation of Addition of Analyzed Rod Position Sequence This proposed change would allow startup sequence modifications beyond those allowed by the general requirements of the BPWS and would minimize unnecessary reactivity manipulations and associated operational challenges. This change would allow control rods to remain inserted in control cells with identified fuel channel deformation. The change would also conform to the existing CRDA bounding assumptions and acceptance criteria.
The licensee performs the CRDA analysis each cycle and documents the thermal operating limit and rod block setpoint results in the core operating limits report in accordance with TS 5.6.5. In its LAR [1], the licensee stated that rod sequence patterns do not fall within the category of information currently specified by technical specifications or the core operating limits report. The licensee uses station processes and procedures for calculations to control the development, approval, and documentation of analyzed control rod sequences consistent with the BPWS methodology. The licensee documents sequences in engineering calculations, and existing administrative controls would continue to provide a back-up methodology to the RWM in assuring compliance with proper rod insertion or withdrawal sequences.
The ANP-10333P-A, Rev. 0 [9], AURORA-B CRDA models can evaluate sequences that are beyond those provided by the BPWS methodology documented in NEDO-21231 [7]. In its LAR [1], the licensee stated that the term analyzed rod position sequence is used to indicate that the sequence, regardless of the use of BPWS, would meet the same CRDA technical requirements as BPWS. The licensee would develop the sequence using the same approved methods as those used to support the current CRDA analysis and implement the sequence in a manner equivalent to those used in the implementation of BPWS-compliant sequences.
In its supplement [2], the licensee provided the NRC staff with the changes to the control rod groupings for the analyzed rod position sequence. This grouping would reassign peripheral rods into groups 5 and 6, which would allow them to be fully withdrawn above the LPSP without having to be positioned at any intermediate notch positions, thereby removing the potential complications of rod settling issues caused by friction. All nonperipheral rods would be assigned to their respective groups in accordance with existing approved BPWS sequences [2]. Further deviations from the methodology not requested in the LAR [1] and its supplement [2], were not reviewed by the NRC staff.
As part of the NRC staff audit [4], the staff observed that the licensee had analyzed the new rod withdrawal sequence and applied the established CRDA methodology [9], to ensure that
7 adequate margin remained to the CRDA bounding acceptance criteria described in UFSAR 15.4.9 [6].
The NRC staff finds that the proposed changes ensure that control rod worths are maintained within the limits prescribed by the postulated reactivity accident for rod dropout during startups and shutdowns, which is described in the UFSAR Section 15.4.9 CRDA, using the existing AURORA-B CRDA methodology [9] approved by the NRC for use at Susquehanna Unit 2 under Amendment 260 [8]. The staff also confirmed that the proposed changes also maintain key safety aspects of the BPWS rules and usage methodology which provides an equivalent level of protection for managing core power distribution during startup and shutdown. Therefore, the NRC staff finds that the proposed changes are acceptable and continue to comply with GDC 28.
3.2 Evaluation of Proposed Changes to TS 3.1.3, 3.1.6, and 3.3.2.1 In its LAR [1], the licensee states:
The proposed change would temporarily modify the current references to Banked Position Withdrawal Sequence (BPWS) and add the analyzed rod position sequence. The analyzed rod position sequence will continue to minimize the consequences of the CRDA. Additionally, the analyzed rod position sequence will provide an equivalent level of protection during plant startups and shutdowns and therefore will not increase the consequences of the CRDA.
As discussed in section 3.1 of this safety evaluation, the staff finds that the proposed analyzed rod position sequence methodology is acceptable. Based on this evaluation, the staff finds that the proposed changes to the LCOs, conditions, required actions, and surveillance requirements listed in section 1.3 of this safety evaluation are adequate, that the changes to the surveillance requirements will assure that the necessary quality of systems and components is maintained and support meeting the LCOs, and that facility operation will be within safety limits because the changes are bounded by the existing CRDA analysis as described in section 15.4.9 of the UFSAR [6]. Therefore, the staff concludes that the proposed changes to the technical specifications comply with the requirements of 10 CFR 50.36(c)(2)(i) and 50.36(c)(3).
3.3 Technical Evaluation Conclusion
The NRC staff finds that the proposed changes conform to the existing NRC-approved CRDA methodology and maintain the current bounding assumptions and acceptance criteria of the UFSAR Section 15.4.9 [6] CRDA. The staff determines that the licensees proposed changes to the rod groupings for the withdrawal sequence are appropriate from a reactivity management standpoint. The staff also confirmed that the proposed changes maintain key safety aspects of the BPWS methodology to the extent practicable. Therefore, the NRC staff finds that the proposed changes are acceptable and maintain compliance with GDC 28.
The NRC staff finds the proposed changes to the LCOs, conditions, required actions, and surveillance requirements in TS 3.1.3, TS 3.1.6, and TS 3.3.2.1 that would temporarily allow implementation of an analyzed rod position sequence and, thus, greater flexibility in rod manipulation during various stages of reactor power operation, are acceptable because the changes are bounded by the existing CRDA analysis as described in section 15.4.9 of the UFSAR [6] and, therefore, meet the requirements of 10 CFR 50.36.
8 4.0 EMERGENCY SITUATION The NRCs regulations in 10 CFR 50.91(a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear powerplant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, the NRC may issue a license amendment involving no significant hazards consideration (NSHC) without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing for opportunity for a hearing and for public comment after issuance.
As discussed in the licensees application dated January 10, 2023, the licensee requested that the NRC process the proposed amendment on an emergency situation basis. In its LAR, the licensee stated that it will be performing a maintenance outage that will require placing the reactor in shutdown. The licensee identified a challenge to comply with the BPWS because of control cell friction caused by fuel channel deformation on high exposure ATRIUM 10 fuel assemblies. This challenge would prevent startup following the maintenance outage without significantly expanding the scope of the maintenance outage (e.g., by developing a withdrawal sequence to comply with BPWS, discharging all high exposure fuel, or rechanneling ATRIUM 10 fuel assemblies).
The NRC staff reviewed the licensees basis for its request and determined that an emergency situation exists consistent with the provisions in 10 CFR 50.91(a)(5). Furthermore, the NRC staff determined that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of 10 CFR 50.91(a)(5). Based on these findings and the determination that the amendment involves NSHC as discussed in section 5.0 of this safety evaluation, the NRC staff has determined that a valid need exists for issuance of the license amendment using the emergency provisions of 10 CFR 50.91(a)(5).
5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves NSHC if operation of the facility, in accordance with the amendment, would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
In its application dated January 10, 2023 [1], the licensee provided its analysis about the issue of NSHC. The licensees supplement [2] provided additional information that clarified the
9 application [1], did not expand the scope of the application, and did not impact the conclusions in the NSHC analysis in the application. The licensees analysis is as follows:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change will modify TS Sections 3.1.3, 3.1.6, and 3.3.2.1.
The proposed change would temporarily modify the current references to Banked Position Withdrawal Sequence (BPWS) and add the analyzed rod position sequence. The analyzed rod position sequence will continue to minimize the consequences of the CRDA. Additionally, the analyzed rod position sequence will provide an equivalent level of protection during plant startups and shutdowns and therefore will not increase the consequences of the CRDA.
Control rod patterns during startup and shutdown conditions will continue to be controlled by the plant operator and the Rod Worth Minimizer (RWM)
(LCO 3.3.2.1), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10 % of Rated Thermal Power (RTP). As a result of this change, these sequences will continue to limit the potential amount of reactivity addition that could occur in the event of a CRDA.
Accidents are initiated by the malfunction of plant equipment, or the failure of plant structures, systems, or components. There are no changes being implemented to plant structures, systems, or components. The proposed changes will ensure that incremental control rod reactivity worths continue to be minimized by implementing rod withdrawal sequences that comply with the analyzed rod position sequence developed in accordance with the NRC approved Framatome Topical Report ANP-10333P-A methodology implemented in Susquehanna TS 5.6. These analyzed rod position sequences will limit the potential reactivity increase for a postulated CRDA during reactor startups and shutdowns below the Low Power Setpoint of 10 % of RTP.
The proposed change will continue to ensure that systems, structures, and components are capable of performing their intended safety functions.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed change does not affect the assumed accident performance of the control rods, nor any plant structure, system or component previously
10 evaluated. The change does not involve a physical alteration of the plant (i.e., no different SSCs [structures, systems, and components] will be installed) or a change in the methods governing normal plant operations. The analyzed rod position sequence will be established pursuant to the approved methods controlling normal plant operations. As such, the proposed change does not introduce new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing basis.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No The proposed change ensures that analyzed rod position sequences are developed to minimize incremental control rod reactivity worth in accordance with the Reference 1 [Reference 1 in the LAR] NRC approved methodology implemented in Susquehanna TS 5.6. Cycle-specific CRDA results are reviewed and approved in accordance with the 10 CFR 50.59 process. The proposed change will not adversely impact the plants response to an accident or transient. All current safety margins will be maintained. There are no changes proposed which alter the set points to which protective actions are initiated and there is no change to the operability requirements for equipment assumed to operate for accident mitigation.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on its review of the licensees evaluation above, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that NSHC is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria in 10 CFR 50.91.
6.0 STATE CONSULTATION
In accordance with the Commissions regulations in 10 CFR 50.91(b), the NRC staff notified the Commonwealth of Pennsylvania officials on January 11, 2023 [10], of the proposed issuance of the amendment. The Commonwealth officials and the NRC staff met via a video conference on January 12, 2023, to discuss the LAR. The Commonwealth officials had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20, Standards for protection against radiation, and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. In section 5.0 of this safety evaluation, the NRC staff has made a final determination that NSHC is involved for the proposed
11 amendment. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
8.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.
9.0 REFERENCES
[1] Jones, D., Susquehanna Nuclear, LLC, letter to U.S. Nuclear Regulatory Commission, Susquehanna Steam Electric Station Proposed Emergency Amendment to Licensee NPF-22: Temporary Addition of Analyzed Rod Position Sequence PLA-8042, January 10, 2023, ADAMS Accession No. ML23010A265.
[2] Jones, D., Susquehanna Nuclear, LLC, letter to U.S. Nuclear Regulatory Commission, Susquehanna Steam Electric Station Response to Request for Additional Information Regarding License Amendment Requesting Temporary Addition of Analyzed Rod Position Sequence PLA-8048, January 14, 2023, ADAMS Accession No. ML23014A002.
[3] Klett, A., U.S. Nuclear Regulatory Commission, letter to Jones, D., Susquehanna Nuclear, LLC, Susquehanna Steam Electric Station, Unit 2 - Regulatory Audit Plan in Support of License Amendment Request to Revise Certain Control Rod Technical Specifications (EPID L-2023-LLA-0003), January 11, 2023, ADAMS Accession No. ML23010A087.
[4] Klett, A., U.S. Nuclear Regulatory Commission, letter to Jones, D., Susquehanna Nuclear, LLC, Susquehanna Steam Electric Station, Unit 2 - Summary of Regulatory Audit in Support of License Amendment Request to Revise Certain Control Rod Technical Specifications (EPID L-2023-LLA-0003), January 15, 2023, ADAMS Accession No. ML23010A088.
[5] Klett, A., U.S. Nuclear Regulatory Commission, email to Brown, K., Susquehanna Nuclear, LLC, NRC Request for Additional Information - Susquehanna License Amendment Request (EPID L-2023-LLA-0003), January 13, 2023, ADAMS Accession No. ML23013A208.
[6] Cimorelli, K., Susquehanna Nuclear, LLC, letter to U.S. Nuclear Regulatory Commission, Susquehanna Steam Electric Station Submittal of Updated Final Safety Analysis Report Revision 70 and Fire Protection Review Report Revision 24, PLA-7935, October 12, 2021, ADAMS Accession No. ML21294A245.
[7] Paone, C.J., General Electric, Licensing Topical Report NEDO-21231, Banked Position Withdrawal Sequence, January 1977, ADAMS Accession No. ML090771242 (nonpublic).
[8] Goetz, S., U.S. Nuclear Regulatory Commission, letter to Cimorelli, K., Susquehanna Nuclear, LLC, Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Nos. 278 and 260 to Allow Application of Advanced Framatome ATRIUM 11 Fuel Methododologies (EPID L-2019-LLA-0153), January 21, 2021, ADAMS Accession No. ML20168B004.
[9] Peters, G., Framatome Inc., letter to U.S. Nuclear Regulatory Commission, Publication of
12 ANP-10333P-A, Revision 0, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA), July 15, 2018, ADAMS Accession No. ML18208A415.
[10] Klett, A., U.S. Nuclear Regulatory Commission, email to Shields, M., Pennsylvania Department of Environmental Protection, NRC Notification to the State of Pennsylvania Re. Susquehanna Steam Electric Station, Unit 2 Emergency Amendment - Control Rod Technical Specifications, January 11, 2023, ADAMS Accession No. ML23012A001.
10.0 ABBREVIATIONS BPWS banked position withdrawal sequence CFR Code of Federal Regulations CRDA control rod drop accident GDC general design criterion LAR license amendment request LCO limiting condition(s) for operation LPSP low power setpoint NRC U.S. Nuclear Regulatory Commission NSHC no significant hazards consideration RTP rated thermal power RWM rod worth minimizer SR surveillance requirement SSC structures, systems, and components TS technical specification(s)
UFSAR updated final safety analysis report 11.0 PRINCIPAL CONTRIBUTORS Charley Peabody, NRR Ravi Grover, NRR Audrey Klett, NRR
ML23010A108 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DEX/EICB/BC NAME AKlett KEntz MWaters (RStattel for)
DATE 01/14/2023 01/14/2023 01/14/2023 OFFICE NRR/DSS/SNSB/BC (A)
NRR/DSS/STSB/BC OGC - NLO NAME DWoodyatt VCusumano MWoods DATE 01/14/2023 01/14/2023 01/14/2023 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME HGonzález AKlett DATE 01/15/2023 01/15/2023