ML20094F833
| ML20094F833 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/09/2020 |
| From: | Robert Kuntz Plant Licensing Branch III |
| To: | Bryan Hanson Exelon Generation Co |
| Kuntz R | |
| References | |
| EPID L-2020-LLA-0063 | |
| Download: ML20094F833 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 9, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)
Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION, UNIT 2 - ISSUANCE OF AMENDMENT NO. 276 RE: INCREASED ALLOWED MAIN STEAM ISOLATION VALVE LEAKAGE (EMERGENCY CIRCUMSTANCES)
(EPID L-2020-LLA-0063 [COVID-19])
Dear Mr. Hanson:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 276 to Renewed Facility Operating License No. DPR-30 for Quad Cities Nuclear Power Station, Unit 2. The amendment consists of changes to the technical specifications (TSs) in response to your application dated April 1, 2020.
The amendment revises the TS 3.6.1.3, Primary Control Isolation Valves (PCIV), Surveillance Requirement 3.6.2.3.10 by revising the combined main steam isolation valve (MSIV) leakage rate limit for all four steam lines from 86 to 93 standard cubic feet per hour (scfh) and the leakage rate through each MSIV leakage path from 34 to 37 scfh. The proposed changes in the allowable limits are a one-time change approved for a single cycle (Cycle 26).
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 due to the time critical nature of the amendment. In this instance, an emergency situation exists due to the Centers for Disease Control recommendations advising social distancing or sequestering staff to prevent the spread of the COVID-19 virus.
A copy of the related safety evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment was issued and the final no significant hazards determination. A Notice of Issuance addressing the final no significant hazards
B. Hanson determination and opportunity for a hearing associated with the emergency circumstances will be included in a future biweekly Federal Register notice.
Sincerely,
/RA/
Robert F. Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-265
Enclosures:
- 1. Amendment No. 276 to DPR-30
- 2. Safety Evaluation cc: Listserv
EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. DPR-30
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by the Exelon Generation Company, LLC (the licensee) dated April 1, 2020 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 to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B. of Renewed Facility Operating License No. DPR-30 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated into this 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 2 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: April 9, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 276 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating areas of change.
REMOVE INSERT License DPR-30 License DPR-30 Page 4 Page 4 Technical Specifications 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 3.6.1.3-8 3.6.1.3-8
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C. The license shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.
D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.
E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1 , which contain Safeguards Information protected under 10 CFR 73.21, is entitled: ?Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2,@ submitted by letter dated May 17, 2006.
Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),
including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 244 and modified by License Amendment No. 254.
F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated July 27, 1979 with supplements dated 1
The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Renewed License No.
DPR-30 Amendment No. 276
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.10 Verify the leakage rate through each MSIV In accordance leakage path is 34 scfh for Unit 1 and with the 37 scfh for Unit 2* when tested at Primary 25 psig, and the combined leakage rate Containment for all MSIV leakage paths is 86 scfh Leakage Rate for Unit 1 and 93 scfh for Unit 2* when Testing Program tested at 25 psig.
- The Unit 2 values are only applicable during Cycle 26.
Quad Cities 1 and 2 3.6.1.3-8 Amendment No. 233/276
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNIT 2 DOCKET NO. 50-265
1.0 INTRODUCTION
By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated April 1, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20092P079), Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs) and renewed facility operating license for the Quad Cities Nuclear Power Station (QCNPS), Unit 2. The proposed changes would revise the QCNPS, Unit 2 TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Surveillance Requirement (SR) 3.6.1.3.10. The proposed change in the allowable limits are a one-time change approved for a single cycle (Cycle 26).
2.0 REGULATORY EVALUATION
2.1 System Description The four main steam lines, which penetrate the drywell, are automatically isolated by the main steam isolation valves (MSIVs). There are two MSIVs on each steam line, one inside containment and one outside containment. The MSIVs are functionally part of the primary containment boundary and leakage through these valves provides a potential leakage path for fission products to bypass secondary containment and enter the environment as a ground level release from the turbine building.
2.2 Description of Requested Change The proposed change modifies the QCNPS, Unit 2, TS 3.6.1.3, SR 3.6.1.3.10 by revising the combined MSIV leakage rate limit for all four steam lines from 86 to 93 standard cubic feet per hour (scfh) and the leakage rate through each MSIV leakage path from 34 to Enclosure 2
37 scfh. The proposed change in the allowable limits are a one-time change approved for a single cycle (Cycle 26).
2.3 Regulatory Requirements and Guidance The regulation under Title 10 of the Code of Federal Regulations (10 CFR)
Paragraph 50.36(a)(1) states, in part: A summary statement of the bases or reasons for such specifications shall also be included in the application, but shall not become part of the technical specifications.
The regulation under 10 CFR 50.36(b) requires that each license authorizing reactor operation include TSs derived from the analyses and evaluation included in the safety analysis report and amendments thereto.
Paragraph 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 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.
The NRC staffs guidance for review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, dated March 2010 (ADAMS Accession No. ML100351425).
The regulation under 10 CFR 50.67, Accident source term, provides a mechanism for licensed power reactors to replace the traditional source term used in the design-basis accident (DBA) radiological consequence analyses with an acceptable alternative source term (AST).
Licensees using the AST are evaluated against the dose criteria specified in 10 CFR 50.67(b)(2):
10 CFR 50.67(b)(2)(i) requires that 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 25 roentgen equivalent man (rem) total effective dose equivalent (TEDE).
10 CFR 50.67(b)(2)(ii) requires that 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 would not receive a radiation dose in excess of 25 rem TEDE during the entire period of its passage.
10 CFR 50.67(b)(2)(iii) requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions, without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.
The construction permits for QCNPS Units 1 and 2 predate the formal issuance of the current Appendix A, General Design Criteria (GDC), to 10 CFR Part 50. During the construction permit licensing process, Units 1 and 2 were evaluated against the Comment Draft of 70 Criteria, which was issued on July 10, 1967. The design bases of each QCNPS unit were reevaluated at the time of initial Final Safety Analysis Report preparation against the draft of the 70 criteria current at the time of operating license application.
As stated in Section 3.1 of the updated final safety analysis report, based on the understanding of the intent of the proposed criteria current at the time of operating license application, QCNPS conforms with the intent of the Atomic Energy Commission GDC for Nuclear Power Plant Construction Permits. As the GDC were finalized, the requirements were placed in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.
The regulation under 10 CFR Part 50, Appendix A, Criterion 19, Control room, requires that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including LOCAs. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE, or its equivalent to any part of the body, for the duration of the accident.
Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000, provides guidance for meeting requirements in 10 CFR 50.67 (ADAMS Accession No. ML003716792).
The NRC staffs guidance for review of AST is in Standard Review Plan (SRP) Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Term, Revision 0, dated July 2000 (ADAMS Accession No. ML003734190).
3.0 TECHNICAL EVALUATION
3.1 MSIV Leakage In the radiological consequence analysis for the DBA loss-of-coolant accident (LOCA), the MSIVs are postulated to leak at a total design leak rate through all four main steam lines of 150 scfh measured at 43.9 pounds per square inch gauge (psig) allowing a maximum leakage of 60 scfh per line. As previously referenced in this safety evaluation, the DBA LOCA radiological analysis was approved by the NRC using the AST methodology in accordance with 10 CFR 50.67 and GDC 19 of Appendix A to 10 CFR Part 50, and the guidance in RG 1.183 Revision 0 and SRP Section 15.0.1.
The proposed one-time change to the TSs MSIV leakage rate limits for a single cycle (Cycle 26), does not propose to make permanent changes to any design input parameters and assumptions in the MSIV leakage model evaluated using the Radionuclide Transport and Removal and Dose Estimation (RADTRAD) computer code. The current calculated post-LOCA doses using the RADTRAD code as 3.83 rem TEDE (control room), 8.47 rem TEDE (exclusion area boundary), and 16.5 rem TEDE (low population zone) continue to be bounding and remain unchanged.
Since the proposed change to the TSs MSIV leakage rate limits do not change the calculated post-LOCA doses and the MSIV leakage model evaluated using the RADTRAD code, the NRC staff finds there is reasonable assurance that (1) 10 CFR 50.67(b)(2) will continue to ensure that the dose criteria are met, and (2) the intent of GDC 19 to Appendix A of 10 CFR Part 50 will continue to be met. Therefore, the proposed revision to SR 3.6.1.3.10 is acceptable.
3.2 MSIV Leakage Conclusion The NRC staff reviewed the proposed TSs change, as it relates to the DBA LOCA radiological analysis, and determined that it meets the dose criteria in 10 CFR 50.67(b)(2) and requirements in GDC 19 to Appendix A of 10 CFR Part 50 because there is no change to the postulated total design MSIV leak rate of 150 scfh and maximum leakage of 60 scfh per line, and because there is no change to the MSIV leakage model evaluated using the RADTRAD code. Additionally, the change to the TSs, as it relates to the MSIV leakage model, was reviewed for consistency with the guidance in RG 1.183, Revision 0 and SRP Section 15.0.1. For the reasons discussed above, the NRC staff finds there is reasonable assurance that (1) the intent of GDC 19 to Appendix A of 10 CFR Part 50 will continue to be met, and (2) the dose criteria in 10 CFR 50.67(b)(2) will continue to be met. Therefore, the proposed revision to SR 3.6.1.3.10 is acceptable.
3.3 TS Revision In the proposed revised TS, the SR 3.6.1.3.10 leakage rate through each MSIV leakage path when tested at greater than or equal to () 25 psig remains at less than or equal to () 34 scfh for QCNPS, Unit 1, and is increased from 34 scfh to 37 scfh for QCNPS, Unit 2, and the combined leakage rate for all MSIV leakage paths when tested at 25 psig remains 86 scfh for QCNPS, Unit 1, and is increased from 86 scfh to 93 scfh for QCNPS, Unit 2.
Additionally, a footnote denoted by an asterisk (*) is added to SR 3.6.1.3.10, which states that the QCNPS, Unit 2 values are only applicable during Cycle 26.
The proposed TSs MSIV leakage rate limits are more restrictive than the MSIV leakage limits used in the postulated LOCA radiological consequences which is described in Section 15.6.5 of the QCNPS Final Safety Analysis Report, as updated, and in NRC issued Amendment Nos. 233 and 229 for QCNPS, Unit 1 and Unit 2, respectively (ADAMS Accession No. ML062070290).
Since the proposed change to the TS MSIV leakage rate limits are derived from the analysis of the postulated LOCA radiological consequences and are more restrictive than those used in the analysis, the NRC staff finds there is reasonable assurance that (1) 10 CFR 50.36(b) will continue to be met, and (2) 10 CFR 50.36(c)(3) will continue to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation are met. Therefore, the proposed revision to SR 3.6.1.3.10 is acceptable.
The regulation at 10 CFR 50.36(a)(1) states, in part: A summary statement of the bases or reasons for such specifications shall also be included in the application, but shall not become part of the technical specifications. Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases changes that correspond to the proposed TS changes for information only. The licensee stated in the license amendment request that it will make supporting changes to the TS Bases in accordance with TS 5.5.10, Technical Specifications (TS) Bases Control Program.
3.4 TS Revision Conclusion The NRC staff reviewed the proposed TS change and determined that it meets the requirements in 10 CFR 50.36(b) because it is derived from the analyses and evaluation included in the safety analysis report, and amendments thereto. Additionally, the change to the TSs was reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.0. The NRC staff reviewed the proposed TSs change
against the regulations and concludes that it meets the requirements of 10 CFR 50.36(b) and 10 CFR 50.36(c)(3), for the reasons discussed above, and thus provides reasonable assurance that the TSs will 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. Therefore, the NRC staff concludes that the proposed TSs change is acceptable.
4.0 EMERGENCY SITUATION
Background
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 power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, it may issue a license amendment involving no significant hazards consideration 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 application dated April 1, 2020, the licensee requested that the proposed amendment be processed by the NRC on an emergency basis. The license amendment request stated that:
During the Main Steam Isolation Valve (MSIV) local leakage rate testing (LLRT) performed on March 30, 2020, marginal results were obtained on an outboard valve that led to the maximum combined leakage path limit of 86 standard cubic feet per hour (scfh) being exceeded. The requested TS amendment supports deferral of repairs until the next Unit 2 refueling outage.
Emergency circumstances are present in that the Centers for Disease Control has issued recommendations advising social distancing or sequestering staff to prevent the spread of the COVID-19 Virus. The nature of the MSIV repair is in conflict with the recommendations in that it requires technicians to be in constant proximity to each other in a hot environment that exponentially increases the likelihood of individuals contracting COVID-19 and potentially inducing a rapid spread. Additionally, the majority of the specialty technicians that are needed to execute this repair are high risk candidates (based on age group) for contracting COVID-19. Losing resources due to a virus spread would cause a situation where the proper technical knowledge would not be available to satisfactorily complete this work (minimal 14 day isolation and likely to be more than one individual based on having to work in close proximity for the work). Additionally, if a seat replacement is needed it would require pre- and post-heat treatment which requires out of state vendors, which may not be able to travel based on individual state restrictions on travel. This leaves the possibility of not being able to properly repair the valve once disassembled due to the likelihood of having to swap specialty technicians out with little to no notice.
QCNPS, Unit 2 is in a planned outage. For the unit to return to power, the MSIV leakage must be within the TS allowed limits. Therefore, failure to approve the amendment in a timely manner could result in a delay in resumption of operations.
4.1 Emergency Situation Conclusion The NRC staff reviewed the licensees basis for processing the proposed amendment as an emergency amendment (as discussed above) and agrees 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 no significant hazards consideration as discussed below, 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 no significant hazards consideration 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.
An evaluation of the issue of no significant hazards consideration is presented below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The increase in the total MSIV leakage rate limit has no impact on the radiological consequence analysis of the Loss of Coolant Accident (LOCA) so the dose consequences of this limiting Design Basis Accident (DBA) remain within the acceptance criteria provided by the NRC for use with the Alternative Source Term (AST) methodology in 10 CFR 50.67 and 10 CFR 50, Appendix A, GDC 19. Additional guidance is provided in Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors and Standard Review Plan (SRP) Section 15.0.1.
The proposed change to the MSIV leakage limit does not involve physical change to any plant structure, system, or component. As a result, no new failure modes of the MSIVs have been introduced.
The proposed change does not affect the normal design or operation of the facility before the accident; rather, it potentially affects leakage limit assumptions that constitute inputs to the evaluation of the consequences.
The radiological consequences of the analyzed LOCA are unchanged from the approved AST doses and continue to have adequate margin to the regulatory limits specified in 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room operator doses is still available.
Therefore, the proposed changes do 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 change in the MSIV leakage rate limits does not affect the design, functional performance, or normal operation of the facility. Similarly, it does not affect the design or operation of any component in the facility such that new equipment failure modes are created. This is supported by operating experience at other EGC [Exelon Generation Company, LLC]
sites that have increased their MSIV leakage limits. As such the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No This proposed license amendment involves changes in the MSIV leakage rate limits. The revised leakage rate limits remain consistent with the LOCA radiological consequence evaluation assumptions. The dose consequences of this limiting event are unchanged by this amendment and remain within the acceptance criteria presented in 10 CFR 50.67 for offsite doses and 10 CFR 50, Appendix A, GDC 19 for control room operator doses. The margin of safety is that provided by meeting the applicable regulatory limits.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on the above evaluation, 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 no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.
6.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendment on April 3, 2020. The State official had the following comment: The request for all the MSIVs doesnt seem to be warranted. The technical justification for one outage cycle for the one MSIV that didnt meet the leakage requirements appears adequate. IEMA [Illinois Emergency Management Agency] cannot comment on whether the licensee meets NRC requirements for this amendment being considered as an
emergency amendment. The NRC staff finds that the total MSIV leakage is based on the contribution of each MSIV and is less than the sum of the leakage allowed for each of the individual valves in the proposed amendment. The NRC staff reviewed the proposed TS change, both with respect to the combined MSIV leakage rate limit and the leakage rate limit for each MSIV, and determined it met applicable regulatory requirements. Therefore, the comments received by the State of Illinois are adequately addressed in the amendment request as documented in this safety evaluation.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20 and a change to 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. 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.
Principal Contributors: K. Bucholtz, NRR J. Bettle, NRR R. Clement, NRR Date of Issuance: April 9, 2020
ML20094F833 *via e-mail OFFICE NRR/DORL/LPL3/PM* NRR/DORL/LPL3/LA* NRR/DSS/SCPB/BC* NRR/DRA/ARCB/BC*
NAME RKuntz SRohrer (JBurkhardt for) BWittick KHsueh DATE 04/03/2020 04/06/2020 04/02/2020 04/03/2020 OFFICE NRR/DSS/STSB/BC* OGC* NRR/DORL/LPL3/BC* NRR/DORL/LPL3/PM NAME VCusumano KGamin NLO NSalgado RKuntz DATE 04/02/2020 04/08/2020 04/09/2020 04/09/2020