ML26022A188
| ML26022A188 | |
| Person / Time | |
|---|---|
| Site: | Columbia  (NPF-021) |
| Issue date: | 02/25/2026 |
| From: | Mahesh Chawla NRC/NRR/DORL/LPL4 |
| To: | Schuetz R Energy Northwest |
| Chawla M, NRR/DORL/LPL4 | |
| References | |
| EPID L-2025-LLA-0057 | |
| Download: ML26022A188 (0) | |
Text
February 25, 2026 Mr. Robert Schuetz Chief Executive Officer Energy Northwest MD 1023 76 North Power Plant Loop P.O. Box 968 Richland, WA 99352
SUBJECT:
COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT NO. 279 TO REVISE TECHNICAL SPECIFICATION 3.3.2.1, CONTROL ROD BLOCK INSTRUMENTATION (EPID L-2025-LLA-0057)
Dear Mr. Schuetz:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 279 to Renewed Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the technical specifications (TSs) in response to your application dated March 21, 2025, as supplemented by letter dated September 29, 2025.
The amendment modifies Columbia TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2, which currently restricts reactor startup with an inoperable rod worth minimizer (RWM) to once per calendar year. The change to the Required Action C.2.1.2 would be to allow additional reactor startups with a new action to verify that control rod coupling inspection has been performed prior to a reactor restart.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. Amendment No. 279 to NPF-21
- 2. Safety Evaluation cc: Listserv
ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 279 License No. NPF-21
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Energy Northwest (the licensee), dated March 21, 2025, as supplemented by letter dated September 29, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and 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 rules and 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 Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-21 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 279 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Mahoney, Acting Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-21 and the Technical Specifications Date of Issuance: February 25, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.02.25 08:32:34 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 279 RENEWED FACILITY OPERATING LICENSE NO. NPF-21 COLUMBIA GENERATING STATION DOCKET NO. 50-397 Replace the following pages of Renewed Facility Operating License No. NPF-21 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specification REMOVE INSERT 3.3.2.1-2 3.3.2.1-2
Renewed License No. NPF-21 Amendment No. 279 (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 279 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.
(3)
Deleted.
(4)
Deleted.
(5)
Deleted.
(6)
Deleted.
(7)
Deleted.
(8)
Deleted.
(9)
Deleted.
(10)
Deleted.
(11)
Deleted.
(12)
Deleted.
(13)
Deleted.
Control Rod Block Instrumentation 3.3.2.1 Columbia Generating Station 3.3.2.1-2 Amendment No. 169 225 226 253 279 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)
C.2.1.2 Verify FRQWUROrod coupling FKHFNVare performed for first 12 rods.
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.
Immediately During control rod movement D. RWM inoperable during reactor shutdown.
D.1 Verify movement of control rods is in compliance with BPWS by a second licensed operator or other qualified member of the technical staff.
During control rod movement 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
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 279 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By application dated March 21, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25083A158), as supplemented by letter dated September 29, 2025 (ML25272A289), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Energy Northwest (or the licensee) submitted a license amendment request (LAR) requesting changes to the technical specifications (TSs) for the Columbia Generating Station (Columbia).
The licensees proposed amendment would modify Columbia TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2, which currently restricts reactor startup with an inoperable rod worth minimizer (RWM) to once per calendar year. The proposed change to the Required Action C.2.1.2 would be to allow additional reactor startups with a new action to verify that control rod coupling inspection has been performed prior to a reactor restart.
The supplemental letter dated September 29, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 10, 2025 (90 FR 24420).
2.0 REGULATORY EVALUATION
2.1
System Description
The RWM system is a non-safety related system. The licensee provided the following system description and operation in the LAR, section 2.1, System Design and Operation:
The Plant Processing Computer (PPC) provides station personnel with real time and historical data used to analyze events and monitor plant performance. The RWM subsystem within the PPC functions as a backup to the Operator during
movement of control rods during reactor startup, shutdown, and low power level (less than 10% rated thermal power) control rod sequences. The use of the RWM minimizes the consequences of a design basis control rod drop accident (RDA) by enforcing pre-established control rod sequences that are designed to maintain the worth of any control rod to acceptable levels as determined by the RDA analysis.
RWM aids in enforcing compliance with the analyzed rod position sequence by initiating appropriate rod withdrawal and rod insertion block signals if the control rod patterns followed by the Operator are not consistent with the pre-established sequence. The pre-established rod position sequences follow the banked position withdrawal sequence (BPWS) analyzed in General Electric Topical Report
[NEDO-21231-A], Banked Position Withdrawal Sequence (BPWS), [Revision 1 (ML25248A319)]. This sequence is stored in the PPC memory and is based on control rod withdrawal procedures designed to limit (and thereby minimize) individual control rod worths such that the peak fuel enthalpy would remain below the specific energy design limit of 280 calories per gram (cal/g) should an RDA occur. Conformance to the design basis enthalpy value ensures that the offsite dose consequences due to an RDA will be within the guidelines of 10 CFR 50.67, Accident Source Term.
The RWM does not interfere with normal reactor operation, does not affect any normal instrumentation displays associated with the selection of a control rod, and does not have any plant control functions. The RWM will not function on a loss of offsite power and the failure of the RWM does not itself cause rod patterns to be established. The RWM function is automatically bypassed at power levels greater than 10% rated thermal power, as there are sufficient void concentrations to preclude a rod drop exceeding 280 cal/g peak fuel enthalpy. At power levels less than 10% rated thermal power, the RWM may be manually bypassed by specific procedural control and verification by a second licensed Operator (or qualified member of the technical staff) that the first Operator is performing control rod movements in accordance with the BPWS.
2.2 Regulations The regulations in 10 CFR 50.36(c)(2)(i) state, in part, Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
The following General Design Criteria (GDC) in 10 CFR 50 Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, are applicable to the review of this LAR:
GDC 20, Protection system functions, states that, The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operation occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
GDC 26, Reactivity control system redundancy and capability, states that, Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
GDC 28, Reactivity limits, states that, 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.
2.3 Precedents While there is no approved precedence to permanently eliminate the limitation of one startup per calendar year with an inoperable RWM, the plants listed below received approval to temporarily lift this restriction.
Oyster Creek Nuclear Generating Station was granted approval in Amendment No. 113 dated November 7, 1986, for a temporary note allowing unlimited startups with an inoperable RWM during their Operating Cycle 11 (ML011160423). In addition to the use of existing procedural controls, Oyster Creek committed to the use of control rod pattern templates when the RWM was inoperable.
James A. FitzPatrick Nuclear Power Plant was granted approval on November 19, 2024, for a temporary note that would not restrict reactor startup with an inoperable RWM while compensatory measures were implemented (ML24313A147).
3.0 TECHNICAL EVALUATION
The subsections below provide the NRC staff evaluation of the impact of the bypass or inoperability of the RWM system and any impact on the related accident analysis.
3.1 Accident Analysis The analysis of the following accidents that may occur during the reactor startup mode may be affected due to inoperability or bypass of the RWM system:
Rod Drop Accident Rod Withdrawal Error Event 3.1.1 Rod Drop Accident As stated in Columbia Final Safety Analysis Report (FSAR) section 15.4.9, Control Rod Drop Accident (ML23346A215), the RDA is the result of a high worth control rod decoupled from the drive mechanism, dropping out of the core. The decoupling of the control rod would insert positive reactivity in the core resulting in a localized power excursion. This is not an anticipated event because of the system failures and personnel errors that would have to occur in combination to present the reactivity required at the same time that the coupling failed. The control rod patterns are controlled in accordance with the BPWS to preclude situations in which rod drops would have sufficient reactivity to cause the damage postulated by the power excursion.
The licensee provided the following reasons why the RWM system is not required to mitigate RDA and thereby prevent conditions in which dropped rods would have sufficient reactivity to cause the damage due by the power excursion:
The control rod withdrawal sequences and patterns during startup are selected prior to operation. Operable control rods are required to comply with the requirements of the BPWS when reactor power is less than 10 percent rated thermal power, to ensure the initial conditions of the RDA analysis are not violated.
The RDA consequences are mitigated by the average power range monitor (APRM),
which generates a high flux scram signal to the reactor protection system, resulting in an automatic scram of the reactor.
Columbia FSAR section 7.7.1.2.2.2. Rod Block Trip System, paragraph b.2.(e) states:
The RWM can initiate a rod insert block and a rod withdrawal block. The purpose of these functions is to reinforce procedural controls that limit the reactivity worth of control rods under lower power conditions. The rod block trip settings are based on the allowable control rod worth limits established for the design basis rod drop accident. Adherence to prescribed control rod patterns is the normal method by which this reactivity restriction is observed.
According to the above description, the RWM system is provided with rod block and rod withdrawal trip settings established for the RDA. The licensee provided the following explanation on how the rod insert block trip and rod withdrawal trip functions are accomplished in the absence of RWM in its supplemental letter dated September 29, 2025.
The terms rod withdrawal block and rod withdrawal trip are used interchangeably, particularly in Final Safety Analysis Report (FSAR)
Section 7.7.1.2.2.2, Rod Block Trip System. The Rod Worth Minimizer (RWM) is
only capable of generating rod insertion blocks and rod withdrawal blocks; the system is not capable of generating a scram signal.
There is no other system or component that generates the rod insertion block or rod withdrawal block at the same settings as the RWM. However, in the absence of the RWM System, second licensed Operator or qualified member of the technical staff enforces adherence to the established startup, shutdown, and low power level control rod sequences. This enforcement is preventive in nature, as the control rod selection is verified prior to rod movement made by the first Operator, preventing the establishment of control rod patterns that are not consistent with the approved rod position sequence. Verification that control rod movement is in compliance with the banked position withdrawal sequence is required to meet Technical Specification 3.1.6, Rod Pattern Control, and Technical Specification 3.3.2.1, Control Rod Block Instrumentation, requirements.
3.1.2 Rod Withdrawal Error Event FSAR section 15.4.1.2, Continuous Rod Withdrawal During Reactor Startup, subsection 15.4.1.2.1, Identification of Causes and Frequency Classification, states:
This event is categorized as an infrequent incident. The probability of further development of this event is low because it is contingent upon the failure of the rod worth minimizer (RWM) system or failure of a second licensed operator (or technically qualified member of the technical staff) observing the out-of-sequence rod selection concurrent with a high worth rod, out-of-sequence rod selection contrary to procedures, and operator disregard of continuous alarm annunciations prior to safety system actuation.
The RWM system is a backup to protect the reactor by preventing operators from moving an incorrect control rod. Therefore, if RWM is inoperable during reactor startup, the RWE event is prevented by a second licensed operator or a qualified member of the technical staff verifying that the control rod movement complies with the BPWS. The use of a second operator or qualified member of the technical staff was found to be acceptable in an NRC safety evaluation for Amendment No. 149 of Columbias TSs (ML17292A736).
3.2 Inoperability or Bypass of RWM System If the RWM is bypassed or inoperable, its block function described in Columbia FSAR section 7.7.1.2.2.2 would be disabled. On this occasion, according to FSAR section 7.7.1.10.2, Operation, the following specific procedural controls are required to be implemented:
- a. Plant management approval is required,
- b. A second operator or technically qualified plant staff member, with no other duties, is required to verify the first operators actions while the first operator is performing rod movements,
- c. The startup and shutdown sequences with their respective signoff sheets are provided to the second operator for verification of each step rod movement made by the first operator, and d.
The startup and shutdown sequences follow the same control rod patterns that the RWM enforces if it were not bypassed.
In the supplement to the LAR dated September 29, 2025, the licensee stated that the proposed change does not affect the above procedural controls and their implementation. However, future changes in these controls would require a license amendment because the controls for items b and d are enforced by the following Required Action C.2.2 of Technical Specification 3.3.2.1, Control Rod Block Instrumentation.
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.
The NRC staff finds the licensees evaluation of an inoperable or bypass of the RWM system acceptable because the licensing basis procedural control remains unaffected and the licensee confirmed that any changes in controls b and d will require a license amendment.
3.3
SUMMARY
3.3.1 Technical Conclusions In case the RWM system is bypassed or inoperable due to any reason, the NRC staff finds the changes to TS 3.3.2.1 acceptable from the accident analysis standpoint and its inoperability or bypass status because of the following reasons:
The operator will maintain acceptable rod worth by adhering to station procedures and prescribed control rod patterns and sequences when the plant is below 10 percent rated thermal power during which the RWM system is intended to be used.
Another layer of defense-in-depth for preventing a design basis RDA with excessive rod worth is provided by TS 3.3.2.1 Condition C.2.2 which requires a second licensed operator or qualified member of the technical staff to verify that the movement of control rods is in compliance with the BPWS.
The inoperability or the bypass status of the RWM system is acceptable because it is not intended to replace operator selection of control patterns but is intended simply to monitor and reinforce procedural adherence.
Coupling checks proposed in TS 3.3.2.1 reduces the probability of an RDA because these checks will ensure the control rod is properly coupled with its drive mechanism in support of a reactor startup.
3.3.2 Regulatory Conclusions The NRC staff has determined that the requirements of 10 CFR 50.36(c)(2)(i) continue to be met because the changes to TS 3.3.2.1 do not change the lowest functional capability or performance levels of equipment required for safe operation of the facility.
The NRC staff concludes that the proposed change meets the requirements of the applicable GDCs for the reasons given below:
GDC 20 continues to be satisfied because the operation of the reactivity control system is not affected ensuring that fuel design limits are not exceeded. The RDA consequences are sensed and is mitigated by the APRM, which generates a high flux scram signal to the reactor protection system, resulting in an automatic scram of the reactor.
GDC 26 and GDC 28 continue to be satisfied because the operation of the reactivity control system is unaffected and the current accident analysis remains valid.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Washington State official was notified of the proposed issuance of the amendment on January 16, 2026. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on June 10, 2025 (90 FR 24420), and there has been no public comment on such finding. 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.
6.0 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.
Principal Contributor: A. Sallman, NRR Date: February 25, 2026
ML26022A188 NRR-058 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DEX/EICB/BC (A)
NAME MChawla PBlechman SDarbali DATE 1/29/2026 1/29/2026 1/29/2026 OFFICE NRR/DSS/SNSB NRR/DSS/STSB NRR/DORL/LPL4/BC (A)
NAME NDifrancesco SMehta MMahoney DATE 1/30/2026 2/9/2026 2/25/2026 OFFICE NRR/DORL/LPL4/PM NAME MChawla DATE 2/25/2026