ML23005A176
| ML23005A176 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/23/2023 |
| From: | V Sreenivas Plant Licensing Branch 1 |
| To: | Rhoades D Constellation Energy Generation |
| Sreenivas V, NRR/DORL/LPL1, 415-2596 | |
| References | |
| EPID L-2022-LLA-0057 | |
| Download: ML23005A176 (1) | |
Text
February 23, 2023 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
R. E. GINNA NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT NO. 153 RE: REVISE TECHNICAL SPECIFICATIONS FOR THE SPENT FUEL POOL CHARCOAL SYSTEM AND TWO TS ADMINISTRATIVE CHANGES (EPID L-2022-LLA-0057)
Dear Mr. Rhoades:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 153 to Renewed Facility Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant in response to your application dated April 18, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22108A248).
The amendment revised the Technical Specification (TS) 3.7.10 "Auxiliary Building Ventilation System" to delete 3.7.10.3 Surveillance Requirements, delete TS 5.5.10(c) Spent Fuel Pool Charcoal Adsorber System for the Ventilation Filter Testing Program, additions to TS 5.6.5 CORE OPERATING LIMITS REPORT (COLR), and an update to TS 5.5.15 Containment Leakage Rate Testing Program.
D. Rhoades A copy of the related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosures:
- 1. Amendment No. 153 to Renewed License No. DPR-18
- 2. Safety Evaluation cc: Listserv
CONSTELLATION GENERATION COMPANY, LLC DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 153 Renewed License No. DPR-18
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Constellation Generation Company, LLC dated April 18, 2022, 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.
Enclosure 1
- 2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 153, are hereby incorporated in the renewed license. Constellation Generation Company LLC shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Richard Richard V. V. Guzman Date: 2023.02.23 13:33:14 Guzman -05'00' 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: February 23, 2023
ATTACHMENT TO LICENSE AMENDMENT NO. 153 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Replace the following page of Renewed Facility Operating License No. DPR-18 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Page Insert Page 4 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain a marginal line indicating the areas of change.
Remove Page Insert Page 3.7.10-1 3.7.10-1 5.5-8 5.5-8 5.5-11 5.5-11 5.6-2 5.6-2
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 153, are hereby incorporated in the renewed license.
Constellation Energy Generation, LLC shall operate the facility in accordance with the Technical Specifications.
(3) Fire Protection Constellation Energy Generation, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensees amendment request dated March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
(a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
- 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
R. E. Ginna Nuclear Power Plant Amendment No. 153
ABVS 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Auxiliary Building Ventilation System (ABVS)
LCO 3.7.10 The ABVS shall be OPERABLE and in operation.
APPLICABILITY: During movement of irradiated fuel assemblies in the Auxiliary Building when one or more fuel assemblies in the Auxiliary Building has decayed < 60 days since being irradiated.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ABVS inoperable. A.1
- NOTE -
LCO 3.0.3 is not applicable.
Suspend movement of Immediately irradiated fuel assemblies in the Auxiliary Building.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Verify ABVS is in operation. In accordance with the Surveillance Frequency Control Program SR 3.7.10.2 Verify ABVS maintains a negative pressure with In accordance with respect to the outside environment at the Auxiliary the Surveillance Building operating floor level. Frequency Control Program R.E. Ginna Nuclear Power Plant 3.7.10-1 Amendment No. 153
Programs and Manuals 5.5
- 2. Demonstrate that an in-place DOP test of the HEPA filter bank shows a penetration and system bypass < 0.05%.
- 3. Demonstrate that an in-place Freon test of the charcoal adsorber bank shows a penetration and system bypass
< 0.05%, when tested under ambient conditions.
- 4. Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows a methyl iodide penetration of less than 1.5% when tested in accordance with ASTM D3803-1989 at a test temperature of 30C (86F), a relative humidity of 95%, and a face velocity of 61 ft/min.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP frequencies.
5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas decay tanks and the quantity of radioactivity contained in waste gas decay tanks. The gaseous radioactivity quantities shall be determined following the methodology in NUREG-0133.
The program shall include:
- a. The limits for concentrations of hydrogen and oxygen in the waste gas decay tanks and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and R.E. Ginna Nuclear Power Plant 5.5-8 Amendment No. 153
Programs and Manuals 5.5
- b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
- c. A required system redundant to the inoperable support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
5.5.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance- Based Option of 10 CFR 50, Appendix J," dated October 2008.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 60 psig.
The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.2% of containment air weight per day.
Leakage Rate acceptance criteria are:
- a. Containment leakage rate acceptance criterion is 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests;
- b. Air lock testing acceptance criteria are:
- 1. For each air lock, overall leakage rate is 0.05 La when tested at Pa, and
- 2. For each door, leakage rate is 0.01 La when tested at Pa.
R.E. Ginna Nuclear Power Plant 5.5-11 Amendment No. 153
Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
The following administrative requirements apply to the COLR:
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
2.1, "Safety Limits (SLs)",
LCO 3.1.1, "SHUTDOWN MARGIN (SDM)",
LCO 3.1.3, "MODERATOR TEMPERATURE COEFFICIENT (MTC)",
LCO 3.1.4, "Rod Group Alignment Limits",
LCO 3.1.5, "Shutdown Bank Insertion Limit",
LCO 3.1.6, "Control Bank Insertion Limits",
LCO 3.1.8 "PHYSICS TEST Exceptions - MODE 2",
LCO 3.2.1, "Heat Flux Hot Channel Factor ( ())",
LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor( )",
LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)",
LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation",
LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits",
LCO 3.4.5, "RCS Loops - MODES 1 8.5% RTP, 2, and 3",
and LCO 3.9.1, "Boron Concentration."
R.E. Ginna Nuclear Power Plant 5.6-2 Amendment No. 153
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 153 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 CONSTELLATION GENERATION COMPANY, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
By application dated April 18, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22108A248), Constellation Energy Generation, LLC (the licensee) requested three changes to the R. E. Ginna Nuclear Power Plant (Ginna) technical specifications (TS).
The proposed changes include:
revisions to the TS for the Auxiliary Building Ventilation System (ABVS) and Ventilation Filter Testing Program (VFTP),
additions to TS 5.6.5 CORE OPERATING LIMITS REPORT (COLR), and update to TS 5.5.15 Containment Leakage Rate Testing Program.
2.0 REGULATORY EVALUATION
Section 50.36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) is the Commission's regulatory requirement that TSs are needed and that TSs are required to include items in five specific categories related to facility operation.
Paragraph 50.36(c)(2)(ii) of 10 CFR, requires that [a] technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the [criteria set forth in 10 CFR 50.36(d)(2)(ii)(A)-(D)]. Paragraph 50.36(c)(3) of 10 CFR requires that TSs include surveillance requirements, which 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.
General Design Criteria (GDC) 18 states in part Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components.
Enclosure 2
GDC 19 requires in part that adequate radiation protection be provided to permit access to and occupancy of the control room under accident conditions and for the duration of the accident without personnel radiation exposures more than 5 rem to the whole body.
GDC 61 requires that fuel storage and handling systems, radioactive waste systems, and other systems that may contain radioactivity be designed to ensure adequate safety under normal and postulated accident conditions and that they be designed with appropriate containment, confinement, and filtering systems.
10 CFR 50.67, Accident source term, requires that (i) an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE), (ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE, and (iii)
Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, (SRP) Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Rev. 0, July 2000 (ML003734190) https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/, provides guidance to the staff for the review of alternative source term amendment requests. Section 15.0.1 states that the NRC reviewer should evaluate the proposed change against the guidance in Regulatory Guide 1.183. The dose acceptance criteria for the fuel handling accident (FHA) are a TEDE of 6.3 rem at the exclusion area boundary (EAB) for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the low population zone (LPZ), and 5 rem in the control room for the duration of the accident.
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes
3.1.1 ABVS and VFTP Revisions These proposed changes will result in the following TS revisions:
Surveillance Requirement (SR) 3.7.10.3, Perform required Spent Fuel Charcoal Absorber System filter testing in accordance with the Ventilation Filter Testing Program (VFTP) will be deleted in its entirety.
The opening paragraph to TS 5.5.10 will be amended from:
A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems and the Spent Fuel Pool (SFP) Charcoal Adsorber System.
to A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems.
Section c SFP Charcoal Adsorber System will be deleted from TS 5.5.10.
3.1.2 Additions to TS 5.6.5 (Administrative)
The following additions will be made to the list of core operating limits which must be documented in the COLR:
LCO [limiting conditions for operations] 3.1.4, Rod Group Alignment Limits, LCO 3.1.8, PHYSICS TEST Exceptions - MODE 2, and LCO 3.4.5, RCS Loops - MODES 1 8.5% RTP, 2, and 3.
3.1.3 TS 5.5.15 Update (Administrative)
The opening paragraph to TS 5.5.15 will be amended from:
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 2-A, Rev. 0, "Industry Guideline for Implementing Performance- Based Option of 10 CFR 50, Appendix J," Revision 2-A, dated October 2008.
to A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 2-A, "Industry
Guideline for Implementing Performance- Based Option of 10 CFR 50, Appendix J," Revision 2-A, dated October 2008.
3.2 NRC Staff Evaluation of TS Changes 3.2.1 ABVS and VFTP Revisions 3.2.1.1 Background/System Operation The ABVS provides clean, filtered, and tempered air to the operating floor of the auxiliary building (AB) and mitigates the effects of fission product activity. The ABVS sweeps air from the AB over the SFP and Decontamination Pit and into the suction of the exhaust system. The air flows through the SFP charcoal filter bank and through multiple fans and high efficiency particulate air filters before discharging out the Intermediate Building roof vent. The ABVS, in accordance with TS 3.7.10, shall be OPERABLE and in operation during movement of irradiated fuel assemblies in the AB when one or more fuel assemblies in the AB has decayed < 60 days since being irradiated. SR 3.7.10.2 requires that the ABVS maintain a negative pressure with respect to the outside environment at the AB operating floor level. SR 3.7.10.3 requires that the SFP charcoal filters comply with TS 5.5.10 (VFTP) for the ABVS to be OPERABLE. TS 5.5.10 contains specific requirements for total air flow rate testing acceptance criteria, in-place penetration and bypass testing acceptance criteria, and laboratory penetration testing acceptance criteria.
The licensee states the proposed change will not make any physical or design changes to the ABVS nor will it change the way in which the ABVS is operated or controlled. The system will continue to exhaust through currently established flow paths and filters during normal operation and continue to maintain a negative pressure to control the flow of airborne radioactivity. During an FHA, the ABVS will no longer be credited for dose reduction by filtering airborne reactivity before discharging out the plant vent due to removal of credit for the SFP charcoal filters.
However, all automatic responses for the ABVS will continue to actuate upon a high radiation alarm and the ABVS will continue (although not credited) to reduce dose following a design basis accident (DBA). All procedural direction involving the ABVS will remain in effect.
3.2.1.2 Dose Evaluation The staff approved the use of the alternate source methodology at the Ginna, with the approval of License Amendment 87 (ML050320491, as amended in ML051230183).
This amendment proposes to remove TS requirements associated with the ABVS charcoal adsorbers and remove credit for filtration in the radiological accident analysis. Therefore, the licensee evaluated the impact of not relying on ABVS filtration in the event a fuel handling accident. A FHA is the only applicable accident since TS 3.7.10 is only applicable when fuel assemblies are being moved in the Auxiliary Building when one or more assemblies in the Auxiliary Building has decayed less than 60 days since being irradiated.
Per Ginnas updated final safety analysis report, section 15.7.3, it is assumed for an FHA that an assembly is dropped from fuel handling equipment onto the top of the racks in the SFP and the assembly is damaged such that there is a release of all radioactive material from the single assembly. The analysis assumes that there is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay prior to fuel offload. In the current licensing basis, the licensee specifies that the most limiting release path is through the AB south roll-up door and includes no credit taken for dose reduction via the ABVS. An alternate
case is analyzed in which the release path is through the plant vent and credit is taken for dose reduction via the ABVS.
The licensee removed the dose reduction credit of the SFP charcoal filters in the case of the plant vent release path, while keeping other assumptions the same as in the current licensing basis. The licensee indicated that the case which assumed a release path out the AB roll-up door was no longer bounding and that the overall dose consequence of an FHA increased. The licensee indicated that there is a small increase in the maximum dose as a result of a FHA in the control room from about 0.8 rem to less than 1.2 rem. The licensee indicates that the dose at the exclusion area boundary (EAB) and low-population zone (LPZ) were not affected.
The staff reviewed the proposed changes in accordance with the requirements of GDC 19, 10 CFR 50.67, and applicable guidance and dose criteria. The staff performed confirmatory calculations using RADTRAD and found the results to be similar to those reported by the applicant, with the dose in the control room, at the EAB, and LPZ, all well below acceptance criteria. As a result, the NRC staff finds that it is not necessary to credit ABVS filtration during a fuel handling accident to meet dose acceptance criteria. Therefore, the NRC staff finds the proposed revisions to be in accordance with applicable requirements and guidance and to be acceptable.
3.2.1.3 GDC Evaluation GDC 18 The licensee states that there will not be any physical or design changes to the ABVS nor will it change the way in which the ABVS is operated or controlled. The system will continue to exhaust through currently established flow paths and filters during normal operation and continue to maintain a negative pressure to control the flow of airborne radioactivity. The ABVS will no longer be credited for dose reduction during a FHA. All automatic responses for the ABVS will continue to actuate upon a high radiation alarm. Additionally, the ABVS will continue to reduce dose following a DBA. The licensee also states that all procedural direction, including testing, involving the ABVS will remain in effect.
Since system operation will not change and the proposed changes are limited to the dose consequences of a FHA and the equipment credited, the staff finds that the requirements of GDC 18 will continue to be met.
GDC 19 As discussed above in section 3.2.1.2, the applicant indicates that the dose to control room personnel will remain well below the 5 rem limit specified in GDC 19. The NRC staff calculated a similar dose to control room operators following a FHA as those indicated by the applicant. As a result, the NRC staff finds that the requirements of GDC 19 will continue to be met.
GDC 61 The NRC staff concurs that the SFP charcoal adsorbers are not required to maintain postulated accident conditions below federal limits. Additionally, there will not be any physical system changes made and the filtering of the airborne radioactivity will continue to occur. The NRC staff finds that adequate safety under normal and postulated accident conditions is ensured because
the system is still designed and operated with appropriate filtering systems; therefore, the NRC staff finds that the fuel storage and handling systems remain in compliance with GDC 61.
3.2.1.4 TS Evaluation As discussed above, the licensee provided sufficient technical justification for removing credit for the SFP charcoal adsorbers in the FHA. The adsorbers no longer meet any of the four criteria of 10 CFR 50.36 for inclusion in the TS, therefore, the staff finds the deletion of section c of TS 5.5.10 and SR 3.7.10.3 appropriate.
3.2.1.5 Conclusion Based on the technical and regulatory evaluations above, the staff finds the revisions to the TS for the ABVS and VFTP acceptable.
3.2.2 Additions to TS 5.6.5 (Administrative)
These proposed additions, currently captured in the site COLR, are an administrative change that comprise three additions to TS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) to ensure TS 5.6.5a matches the current Ginna COLR.
The staff finds adding the three LCOs listed in Section 3.1.2 of this evaluation to TS 5.6.5a acceptable because it aligns the TS with the current site COLR for Cycle 43, Revision 0.
3.2.3 TS 5.5.15 Update (Administrative)
Currently, TS 5.5.15 references both Rev. 0 and Rev. 2-A for NEI 94-01. When the licensee submitted a previous amendment request (ML19045A282), the associated TS markup did not remove the reference to Rev. 0. However, the evaluation section correctly identified the need to remove Rev. 0. The correct revision, Rev. 2-A, was part of the associated approved amendment 136 (ML19325D824).
The staff finds this administrative correction acceptable because it aligns the TS text with a previously approved amendment which reflects the documents accurate revision number.
4.0 TECHNICAL CONCLUSION Based on the technical evaluation provided above, the NRC staff finds the changes adhere to the regulations contained in 10 CFR 50.36. Additionally, the dose evaluation ensures that there is reasonable assurance that there is adequate protection of the public and the facilitys operators. Therefore, the NRC staff finds that the proposed changes are acceptable.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the staff notified the New York State official on January 4, 2023, of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 or change inspections or 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. The Commission has previously issued a proposed finding, which was published in the Federal Register on June 14, 2022 (87 FR 36009) that the amendment involves no significant hazards consideration, 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) and (10). 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: D Scully, NRR E Stutzcage, NRR Date: February 23, 2023
ML23005A176 *by memorandum OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/STSB/BC NAME VSreenivas KZeleznock VCusumano DATE 01/04/2023 01/23/2023 11/30/2022 OFFICE OGC/NLO NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME KDowling HGonzález (RGuzman for) VSreenivas DATE 02/01/2023 02/23/2023 02/23/2023