ML24313A147
| ML24313A147 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/19/2024 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| EPID L-2024-LLA-0134 | |
| Download: ML24313A147 (16) | |
Text
November 19, 2024 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT ISSUANCE OF AMENDMENT NO. 358 RE: REVISE TECHNICAL SPECIFICATIONS TO PERMIT REACTOR STARTUP WITH THE ROD WORTH MINIMIZER INOPERABLE (EPID L-2024-LLA-0134)
Dear David Rhoades:
The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 358 to Renewed Facility Operating License No. DPR-59 for the James A.
FitzPatrick Nuclear Power Plant. The amendment consists of changes to the technical specifications (TSs) in response to the application from Constellation Energy Generation, LLC dated September 25, 2024, as supplemented by letter dated October 29, 2024.
This amendment revises TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2 to temporarily permit reactor startup with the rod worth minimizer inoperable while compensatory measures are implemented. This temporary allowance expires December 31, 2024.
A copy of the NRCs related safety evaluation is also enclosed. A notice of issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosures:
- 1. Amendment No. 358 to DPR-59
- 2. Safety Evaluation cc: Listserv
CONSTELLATION FITZPATRICK, LLC AND CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 358 Renewed Facility Operating License No. DPR-59
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation FitzPatrick, LLC and Constellation Energy Generation, LLC (collectively, the licensees) dated September 25, 2024, as supplemented on October 29, 2024, 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. DPR-59 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 358, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 3 business days.
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: November 19, 2024 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2024.11.19 10:08:44 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 358 JAMES A. FITZPATRICK NUCLEAR POWER PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 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 Page Insert Page 3.3.2.1-2 3.3.2.1-2 Amendment 358 Renewed License No. DPR-59 (3)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus, components or tools.
(5)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30 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 renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Constellation Energy Generation, LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 2536 megawatts (thermal).
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 358, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
Control Rod Block Instrumentation 3.3.2.1 JAFNPP 3.3.2.1-2 Amendment 358 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
(continued)
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 current 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.
Immediately 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 (continued)
- Reactor startup with the RWM inoperable is permitted while the compensatory measure described in letter JAFP-24-0047 dated September 25, 2024, is implemented. This allowance expires on 12/31/2024 at 23:59.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 358 CONSTELLATION FITZPATRICK, LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59
1.0 INTRODUCTION
1.1 Background
By letter dated September 25, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24269A132), as supplemented by letter dated October 29, 2024 (ML24303A408), Constellation Energy Generation, LLC (Constellation, the licensee) submitted a request for changes to the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) Technical Specifications (TSs). The supplemental letter dated October 29, 2024 (ML24303A408), 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) staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 3, 2024 (89 FR 80607).
The proposed changes would revise TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2 to temporarily permit reactor startups in calendar year (CY) 2024 with the rod worth minimizer (RWM) inoperable, and compensatory measures implemented. The licensee is experiencing problems with the RWM software, which is blocking control rod movements that should be allowed in accordance with the reactivity management plan. As a result, the licensee declared the RWM inoperable on September 17, 2024, for the reactor startup following a planned refueling outage. On September 23, 2024, the licensee experienced an automatic trip of the generator output breakers, which caused a reactor scram following a main turbine trip.
The licensee requested, and the NRC granted, a notice of enforcement discretion (NOED) on September 26, 2024 (ML24270A145, verbally granted on September 23, 2024), whereby the NRC exercised discretion to not enforce compliance with TS limiting condition for operation (LCO) 3.3.2.1. for one additional reactor startup in CY 2024 between September 23, 2024, and September 25, 2024, with compensatory measures in place.
FitzPatrick commenced reactor startup on September 24, 2024, with the RWM operable and enhanced procedural guidance in place. Reactor startup with the RWM operable continued up to and beyond 10% reactor power, which is beyond the point where the RWM is required to be operable. FitzPatrick remained compliant with TS LCO 3.3.2.1 and the requested and subsequently approved NOED was not utilized.
1.2 Description of System
Section 7.16 of the FitzPatrick Updated Final Safety Analysis Report (UFSAR) (ML23219A110),
describes the emergency and plant information computer and RWM functions. The RWM system is part of the overall plant process computer system, which monitors and logs different process inputs and makes analytical computations. Specifically, the RWM exists to supplement operator controls by preventing control rod withdrawal if the selected control rod is not part of the preplanned withdrawal sequence during low power operations. These controls are used to ensure reactivity management through normal operations and anticipated operational occurrences does not exceed existing analysis. The RWM will initiate appropriate rod withdrawal block and rod insert block interlock signals to the reactor manual control system rod block circuitry if the control room operator attempts to deviate from pre-set control rod patterns.
These patterns are stored as RWM sequences loaded in the computer memory and are based on control rod withdrawal procedures designed to limit control rod worth to acceptable levels.
The RWM provides redundant reactivity management controls to mitigate against a control rod drop accident and an incorrect rod selection which would lead to a rod withdrawal error.
As indicated in the UFSAR and the license amendment request (LAR), the RWM is used during reactor startup, shutdown, and low-power (less than ten percent) operation to act as a supplemental control to ensure the fuel remains within its design basis. The RWM computer initiates rod withdrawal blocks and rod insert block interlocks to assist the control room operator in selection of control rods in accordance with the banked position withdrawal sequence (BPWS). This predetermined sequence of control rod withdrawals during startup ensures acceptably low individual rod worth to minimize the consequences of a reactivity accident. Compliance with the BPWS is enforced with the RWM and operator oversight of rod selection.
1.3 Proposed Change The licensee proposed to change TS 3.3.2.1, Condition C for the inoperability of the RWM during reactor startup. The change would add a footnote to TS 3.3.2.1, Required Action C.2.1.2, to permit reactor startup with the RWM inoperable using compensatory measures. This allowance would expire on December 31, 2024.
Currently, TS 3.3.2.1, Condition C only permits a single reactor startup per CY with an inoperable RWM. The licensees proposed amendment would allow for additional time to troubleshoot, test, and repair faulty RWM software while not restricting any potential future reactor startups for the remainder of CY 2024.
2.0 REGULATORY EVALUATION
2.1 Regulations Under Title 10 of the Code of Federal Regulations (10 CFR) 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a) (regarding, among other things, consideration of the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals) 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 Commission's regulations.
The regulations in 10 CFR 50.36(c)(2)(i) state, in part, that TS are required to include items in the category of limiting conditions of operation (LCO), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee must shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
2.2 General Design Criteria The construction permit for FitzPatrick was issued by the Atomic Energy Commission (AEC) on May 20, 1970, and the operating license was issued on October 17, 1974. The plant design criteria for the construction phase are listed in the UFSAR, Chapter 1.5, Principal Design Criteria. The AEC published the final rule that added Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program, dated September 18, 1992 (ML003763736), the Commission decided not to apply the final General Design Criteria (GDC) to plants with construction permits issued prior to May 21, 1971, which includes FitzPatrick.
However, the FitzPatrick UFSAR, Chapter 16.6, Conformance to AEC Design Criteria, evaluates FitzPatrick against the 10 CFR 50 Appendix A GDC. Also, the initial AEC safety evaluation of FitzPatrick, dated November 20, 1972 (ML19182A200), Chapter 14.0, stated:
Based on our evaluation of the design and design criteria for the James A.
FitzPatrick Nuclear Power Plant, we conclude that there is reasonable assurance that the intent of the General Design Criteria for Nuclear Power Plants, published in the Federal Register on May 21, 1971[,] as Appendix A to 10 CFR Part 50, will be met.
Therefore, the NRC staff reviews proposed amendments to the FitzPatrick renewed facility operating license using the GDC in 10 CFR Part 50, Appendix A unless there are specific criteria identified in the UFSAR.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition Reactor (the Standard Review Plan (SRP)), Section 4.3, Revision 3, Nuclear Design (ML070740003),Section II, Acceptance Criteria, provides acceptance criteria for reviewing the nuclear design of the fuel assemblies, control systems, and reactor core, which include the following GDCs:
GDC 10 as it relates to designing the reactor core and associated coolant, control, and protection systems with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).
GDC 26 as it relates to the requirement that two independent reactivity control systems of different design be provided, each with the capability to control the rate of reactivity changes resulting from planned, normal power changes.
GDC 28 as it relates to designing the reactivity control systems 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.
Section II, Acceptance Criteria, of Section 15.4.1, Revision 3, Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition (ML063600413), of the SRP provides acceptance criteria for the reviewing the effects and consequences of an uncontrolled control rod assembly withdrawal from a subcritical or low-power condition, which includes the following GDC:
GDC 20 as it relates to the requirement that the protective system automatically initiate the operation of the reactivity control system to ensure that fuel design limits are not exceeded as a result of anticipated operational occurrences to ensure cladding integrity is not challenged.
3.0 TECHNICAL EVALUATION
The NRC staffs evaluation of the licensees analysis of the events that may be affected by the proposed change is given below.
The LAR discusses two main categories of incorrect rod movement which could be impacted with an inoperable RWM: the control rod drop accident and rod withdrawal errors.
3.1 Control Rod Drop Accident The control rod drop accident is defined in chapter 14 of the UFSAR as the assumed drop of the highest worth rod that can be developed at any time in core life or plant operating conditions by one error on the part of the operator.
In the LAR, the licensee evaluates the control rod drop accident at less than ten percent power, when the RWM is required to be operable, and states:
In considering the possibilities of a control rod drop accident, only the rod worths of the lower curve of UFSAR Figure 14.6-1 are pertinent at less than 10 percent power. These are the rods which are normally allowed to be moved by operating procedures and the RWM. The nonscheduled rods, those within the central envelope, do not have a withdrawal permissive during the time their worths are greater than the lower curve, so they are held full in by the control rod drive and cannot drop from the core. If a nonscheduled rod were selected, the RWM blocks rod movement. Therefore, the worth of the strongest rod which could be stuck is limited to about 0.01 k [differential control rod worth]. The 0.025 k worth assumed for cases A and B is considerably above the rod worth values available for stuck rods under the assumed reactor conditions.
Although the RWM has operability challenges, the licensee confirmed in the October 29, 2024, supplement that the challenges are associated with blocking control rod motion when the movement is allowed by BPWS, or by a failure of the system to re-initialize. The challenges have not been associated with failure to block control rod movement when required.
Additionally, the licensee stated in its application that the initial conditions prior to the control rod drop accident assume a control rod pattern in compliance with the BPWS, which limits control rod worth to 1.2% k. This is lower than the initial value in the control rod drop accident analysis, which is assumed to be 2.5% k.
The NRC staff concludes that the proposed change to allow startups for the rest of CY 2024 without an operable RWM will have no impact on the existing control rod drop accident analysis because the BPWS, as enforced by operator oversight and the proposed compensatory measure, limits control rod worth to 1.2% k which is lower than, and bounded by, the 2.5%
k initial value assumed in the plants control rod drop accident analysis. GDC 10, 26, and 28 continue to be met because the existing accident analysis remains valid.
3.2 Control Rod Withdrawal Error In the LAR, the licensee states:
Control rod withdrawal errors are considered when the reactor is at power levels below the power range. The most severe case occurs when the reactor is just critical at room temperature and an out-of-sequence rod is continuously withdrawn. Procedural controls supplemented by the RWM would normally prevent withdrawal of such a rod. It is assumed that the Intermediate Range Neutron Monitoring (IRM) channels are in the worst conditions of allowed bypass.
The scaling arrangement of the IRMs is such that for unbypassed IRM channels a SCRAM signal is generated before the detected neutron flux has increased by more than a factor of ten. In addition, a high neutron flux SCRAM is generated by the average power range monitors.
In the supplement dated October 29, 2024, the licensee confirmed the IRM high power rod block is set to protect against the reactivity excursion created by a control rod withdrawal error.
In the LAR supplement, the licensee states:
If the IRM high power rod block fails to arrest the associated power excursion, the transient is terminated by the IRM scram. As described by Technical Specification (TS)
Bases 3.3.2.1, below 30% power, the consequences of an RWE event will not exceed the safety limit minimum critical power ratio (SLMCPR). Therefore, an inoperable RWM will not result in violation of the SLMCPR or damage to the fuel.
This IRM block is followed by an IRM scram which shuts down the reactor if the rod withdrawal transient is not stopped by the IRM block. Based on this redundant protection, the NRC staff concludes that the fuel design limits are not exceeded even in the worst-case rod withdrawal error.
The NRC staff concludes that the proposed change to allow startups for the rest of CY 2024 without an operable RWM is acceptable based on the review of the worst-case rod withdrawal error. GDC 20 continues to be met because the proposed change does not impact operation of the reactivity control system to ensure that fuel design limits are not exceeded.
3.3 Compensatory Measures In its LAR, the licensee has proposed the following compensatory measure for any reactor startup during the time where this proposed change is applicable: JAF [James A. FitzPatrick]
will dedicate an independent third-party review of the rod movement sheets.
A third party review of the rod movement sheets prior to startup, will provide defense in depth to ensure compliance with the BPWS. This is meant to make up for the loss of redundancy provided by the RWM. As discussed in section 7.16.5.3 of the UFSAR, the RWM function prevents the control room operator from establishing control rod patterns that are not consistent with the pre-planned withdrawal sequences by initiating appropriate rod blocks. In a reactor startup, there are two potential concerns that could lead to a challenge to the fuel integrity: an incorrectly designed rod movement sheet or an incorrectly selected rod in the planned startup.
The first concern is addressed by the compensatory measure to have a third-party review of the rod movement sheets, which ensures the pre-planned sequence has been developed in accordance with the appropriate reactor engineering design requirements. The second concern is mitigated by additional existing TS and station requirements.
If Condition C of TS 3.3.2.1 is entered, Action C.2.2 already requires a second licensed operator or other qualified member of the technical staff to verify movement of control rods is in compliance with BPWS. Furthermore, existing station procedures require a second licensed operator to perform all Main Control Room peer checks and oversight of control rod movement by a dedicated reactivity management senior reactor operator. Although this TS Action followed during all reactor startups as required, not just ones without an operable RWM, the NRC staff finds that they represent a diverse and redundant approach to reactivity management.
The licensee proposed the following footnote under TS 3.3.2.1:
Reactor startup with the RWM inoperable is permitted while the compensatory measure described in letter JAFP-24-0047 dated September 25, 2024, is implemented. This allowance expires on 12/31/2024 at 23:59.
The NRC staff concludes that the licensees proposed compensatory action is sufficient to provide a similar redundancy to the function of the RWM for the limited duration this proposed change is applicable. Staffs conclusion is supported by the existing TS Action and existing station procedures which further reduces the likelihood that the individual control rod worths exceed the acceptable levels. This TS change does not change the lowest functional capability or performance levels of equipment required for safe operation of the facility, but instead provides temporary permission to allow additional startups with the RWM inoperable through the end of the calendar year subject to a compensatory measure. Therefore, the NRC staff finds that the proposed change to the TS with the inclusion of the compensatory measure, is acceptable because the TS, as modified, will continue to meet 10 CFR 50.36(c)(2)(i).
3.4 Risk Insights The licensee included risk insights in the LAR. The licensee stated that the RWM is not modeled in the FitzPatrick probabilistic risk assessment (PRA) because, if a reactivity excursion due to movement of an incorrect control rod were to occur, the reactor SCRAM signal would protect the core from severe accident conditions. The licensee also stated the reactor SCRAM function is modeled in the FitzPatrick PRA using the reactor pressure and water level signals such that if those signals fail, an anticipated transient without SCRAM would occur. The licensee further stated that while the RWM can prevent reactivity excursions and fuel challenges, it provides no additional protection against severe accidents and is of negligible risk significance to public health and safety. The NRC staff reviewed the risk insights provided in the LAR and concluded that the available risk insights support the engineering conclusions provided in this safety evaluation.
3.5 Technical Evaluation Conclusion
The NRC staff concludes that the proposed change to TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2 to permit reactor startup with the RWM inoperable for the remainder of CY 2024 is acceptable based on the following:
The proposed change to allow startups for the rest of the CY without an operable RWM will have no impact on the existing control rod drop accident analysis because the BPWS as enforced by operator oversight and the proposed compensatory measure, taking into account the other compensatory measures already required, limits the values for rod worth to less than is assumed in the analysis.
The proposed change to allow startups for the rest of the CY without an operable RWM is acceptable based on the review of the worst-case rod withdrawal error.
The licensees proposed compensatory action, taking into account the other compensatory actions already required, are sufficient to provide a similar redundancy to the function of the RWM for the limited duration this amendment is applicable.
GDC 10, 26, and 28 continue to be met because the existing accident analysis remains valid.
GDC 20 continues to be met because the proposed change does not impact operation of the reactivity control system to ensure that fuel design limits are not exceeded.
The requirements of 10 CFR 50.36(c)(2)(i) continue to be met because this additional language does not change the lowest functional capability or performance levels of equipment required for safe operation of the facility.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION 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 no significant hazards consideration (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 LAR, the licensee provided its analysis that the proposed amendments involve NSHC. The licensees supplement provided additional information that clarified the application, did not expand the scope of the application, and did not affect the analysis that the proposed amendments involve NSHC. The licensees evaluation of the issue of NSHC is presented below:
- 1. Do the proposed changes involve a significant increase in the probability, or consequences of an accident previously evaluated?
Response: No The proposed RWM bypass allowance does not involve the modification of any plant equipment or affect basic plant operation. The proposed allowances provide additional time to correct problems associated with the RWM. In the event the RWM is inoperable during reactor startup, the technical specification ensures that a licensed operator or other qualified member of the technical staff enforce compliance with the control rod position sequence developed using the banked position withdrawal sequence (BPWS). Applicable compensatory measures will be implemented in the event the RWM is inoperable.
The proposed change does not involve a change to the safety function of the RWM. The proposed change involves no significant changes to the operations of any systems or component in normal or accident operating conditions.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed RWM bypass allowance is not a precursor to any accident previously evaluated. The proposed change provides additional time to rectify RWM equipment issues to ensure that the system implements the control rod pattern developed using BPWS methodology. The proposed change is not required to mitigate the accident conditions. The proposed change does not change the safety function. There is no alteration to the parameters within which the plant is normally operated. The RWM bypass allowance for additional startups is not a precursor to a new or different kind of accident and does not initiate new or different kinds of accidents. As a result, no new failure modes are being introduced.
Therefore, the proposed amendment will not create the possibility of a new or different accident from any accident previously evaluated.
- 3. Do the proposed changes involve a significant reduction in a margin of safety?
Response: No The margin of safety is established through the design of the plant structures, systems, and components, and administrative controls within which the plant is operated. The margin of safety to the consequences of a control rod drop accident is maintained through the use of additional administrative controls described within the current technical specification. The establishment for the control rod insertion and withdrawal during the startups is manually controlled with independent verification by a second licensed reactor operator or other qualified member of the technical staff to ensure compliance with BPWS, if RWM becomes inoperable for any reason. Therefore, the proposed change does not impact the design basis accidents. The proposed change does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.
Therefore, the licensee stated that the proposed amendment does not involve a significant reduction in a margin of safety.
Based on NRCs review of the licensees evaluation, the NRC staff finds that the three standards of 10 CFR 50.92(c) are satisfied. Also, no comments have been submitted on the proposed no significant hazards consideration determination. 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 contained in 10 CFR 50.91.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on November 5, 2024. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to 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, and there has been no public comment on such finding (October 3, 2024; 89 FR 80607). Also, the NRC staff has made a final no significant hazards consideration finding, as discussed above. 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.
7.0 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 Contributors: Jo Ambrosini, NRR Michelle Kichline, NRR
ML24313A147 OFFICE NRR/DORL/LPL1/PMiT NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NAME RLantigua RGuzman KZeleznock DATE 11/08/2024 11/08/2024 11/12/2024 OFFICE NRR/DSS/SNSB/BC NRR/DRA/APOB/BC NRR/DSS/STSB/BC NAME PSahd (RBeaton for)
AZoulis SMehta (KWest for)
DATE 11/04/2024 11/08/2024 11/06/2024 OFFICE OGC (NLO)
NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME KBernstein H González RGuzman DATE 11/15/2024 11/19/2024 11/19/2024