Source: https://www.federalregister.gov/documents/2018/11/20/2018-24894/biweekly-notice-applications-and-amendments-to-facility-operating-licenses-and-combined-licenses
Timestamp: 2019-08-19 01:34:16
Document Index: 120325307

Matched Legal Cases: ['art 2', 'art 50', 'art 50', 'art 50', 'art 0', 'art 50', 'art 50', 'art 50', '§\u2009100', 'art 52']

A Notice by the Nuclear Regulatory Commission on 11/20/2018
Comments must be filed by December 20, 2018. A request for a hearing must be filed by January 22, 2019.
58607-58626 (20 pages)
II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination.
Exelon Generation Company (EGC), LLC, Docket No. 50-461, Clinton Power Station (CPS), Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC (Exelon), Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Southern Nuclear Operating Company, Inc., Docket No. 52-025, Vogtle Electric Generating Plant (VEGP), Unit 3, Burke County, Georgia
Northern States Power Company—Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant (Monticello), Wright County, Minnesota
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle), Burke County, Georgia
https://www.federalregister.gov/d/2018-24894 https://www.federalregister.gov/d/2018-24894
This biweekly notice includes all notices of amendments issued, or proposed to be issued, from October 23, 2018, to November 5, 2018. The last biweekly notice was published on November 6, 2018.
Please refer to Docket ID NRC-2018-0266, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0266.
NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​Start Printed Page 58608adams.html. To begin the search, select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
Please include Docket ID NRC-2018-0266, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.
If a hearing is requested, and the Commission has not made a final Start Printed Page 58609determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.
Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited Start Printed Page 58610delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.
Date of amendment request: July 19, 2018. A publicly-available version is in ADAMS under Accession No. ML18200A252.
Description of amendment request: The amendments would modify the Catawba Updated Final Safety Analysis Report (UFSAR), Section 6.2.4.2.2, “Containment Valve Injection Water System [CVIWS],” to remove the CVIWS supply from specified Safety Injection (NI) and Containment Spray (NS) Containment Isolation Valves (CIVs), and to exempt these CIVs from Type C Local Leak Rate Testing (LLRT). Additionally, the amendments would modify UFSAR, Table 6-77, “Containment Isolation Valve Data,” to make corresponding changes.
The amendment request is to remove select Containment Isolation Valves from the Local Leak Rate Test (LLRT) program. These valves were originally included in the LLRT under 10 CFR 50, Appendix J, in what is now Option A. [Catawba] has been approved for 10 CFR 50, Appendix J, Option B under License Amendment No. 192/184. Under Option B, valves may be exempted from LLRT Type C testing if they are not a potential containment atmosphere leakage path. Based on the design and operation of the NI and NS Systems, the valves do not constitute a containment atmospheric leakage path as covered in the Safety Evaluation. Since the valves are not a leakage path, there is no impact on the consequence of an accident. Moreover, the valves are not a part of the Reactor Coolant Pressure Boundary, thus they do not affect the probability of an accident.
The systems design and operation are not changing. This test exemption does not change the way the valves are used as a part of the NI and NS Systems. A detailed Failure Modes and Effects Analysis was completed to confirm the system operation would meet the containment isolation design function.
The test exemption is within existing regulatory requirements. The application of a closed loop outside of containment is appropriate and consistent with regulatory positions. With containment integrity maintained within the allowable regulatory framework, there is no reduction in the margin of safety.
Date of amendment request: October 2, 2018. A publicly-available version is in ADAMS under Accession No. ML18275A060.
Description of amendment request: The amendment would modify the Technical Specifications concerning a change to the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance at FitzPatrick.
The proposed Technical Specification change does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The amendment would only change how the reactivity anomaly surveillance is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the Technical Specification requirements such that, rather than performing the surveillance by comparing predicted to actual control rod density, the surveillance is performed by a direct comparison of keff. Present day online core monitoring systems, such as the one in use at the James A. FitzPatrick Nuclear Power Plant [(JAFNPP)], Unit 1 are capable of performing the direct measurement of reactivity.
Therefore, since the reactivity anomaly surveillance will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of a previously evaluated accident.Start Printed Page 58611
This Technical Specifications amendment request does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the reactivity anomaly Technical Specifications will only provide a new, more efficient method of detecting an unexpected change in core reactivity.
Date of amendment request: September 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18271A217.
Description of amendment request: The amendment would make Technical Specification (TS) changes that are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-476, Revision 1. The availability of this TS improvement was announced in the Federal Register on May 23, 2007 (72 FR 29004).
The proposed change modifies the TS to allow the use of the improved BPWS [Banked Position Withdrawal Sequence] during shutdowns if the conditions of NEDO-33091-A, Revision 2, “Improved BPWS Control Rod Insertion Process,” July 2004 [ADAMS Accession No. ML042230366], have been satisfied. The justifications to support the specific TS changes are consistent with the approved topical report and TSTF-476, Revision 1. Since the change only involves changes in control rod sequencing, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting TSTF-476 are no different than the consequences of an accident prior to adopting TSTF-476. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change, TSTF-476, Revision 1, incorporates the improved BPWS, previously approved in NEDO-33091-A, into the CPS TS. The CRDA is the design basis accident for the subject TS changes. In order to minimize the impact of a CRDA, the BPWS process was developed to minimize control rod reactivity worth for boiling water reactor plants. The proposed improved BPWS further simplifies the shutdown control rod insertion process, and in order to evaluate it, the NRC followed the guidelines of Standard Review Plan Section 15.4.9, and referred to General Design Criterion 28 of Appendix A to 10 CFR part 50 as its regulatory requirement. The TSTF stated the improved BPWS provides the following benefits: (1) Allows the plant to reach the all-rods-in condition prior to significant reactor cool down, which reduces the potential for recriticality as the reactor cools down; (2) reduces the potential for an operator reactivity control error by reducing the total number of control rod manipulations; (3) minimizes the need for manual scrams during plant shutdowns, resulting in less wear on control rod drive (CRD) system components and CRD mechanisms; and (4) eliminates unnecessary control rod manipulations at low power, resulting in less wear on reactor manual control and CRD system components. The addition of procedural requirements and verifications specified in NEDO-33091-A, along with the proper use of the BPWS will prevent a CRDA from occurring while power is below the low power setpoint (LPSP). The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.
Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, 4300 Winfield Road, Warrenville, IL 60555.
Date of amendment request: July 25, 2018. A publicly-available version is in ADAMS under Accession No. ML18206A545.
Description of amendment request: The amendment would revise the TMI-1 Renewed Facility Operating License (RFOL) and associated Technical Specifications (TSs) to the Permanently Defueled Technical Specifications (PDTSs), consistent with the permanent cessation of reactor operation and permanent defueling of the reactor. By letter dated June 20, 2017 (ADAMS Accession No. ML17171A151), Exelon provided formal notification to the NRC of Exelon's contingent determination to permanently cease operations at TMI-1 no later than September 30, 2019. The amendment would eliminate those TSs applicable in operating mode or modes where fuel is placed in the reactor vessel. The amendment would change other TS limiting conditions for operation (LCOs), definitions, surveillance requirements, and administrative controls, as well as several license conditions. The Start Printed Page 58612amendment would also modify the licensing basis mitigation strategies for flood mitigation and aircraft impact protection in the air intake tunnel.
The proposed changes would not take effect until TMI has certified to the NRC that it has permanently ceased operation and entered a permanently defueled condition. Because the 10 CFR part 50 license for TMI will no longer authorize operation of the reactor, or emplacement or retention of fuel into the reactor vessel with the certifications required by 10 CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 0.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible.
The remaining UFSAR [Updated Final Safety Analysis Report] Chapter 14 postulated design basis accident (DBA) events that could potentially occur at a permanently defueled facility would be a Fuel Handling Accident (FHA) in the Spent Fuel pool (SFP), Waste Gas Tank Rupture (WGTR), and Fuel Cask Drop Accident (FCDA). The FHA analyses for TMI shows that, following 60 days of decay time after reactor shutdown and provided the SFP water level requirements of proposed TS LCO 3/4.1.1 are met, the dose consequences are acceptable without relying on SSCs [structures, systems, and components] to remain functional for accident mitigation during and following the event. The one exception to this is the continued function of the passive SFP structure. The remaining DBAs that support permanently shutdown and defueled condition do not rely on any active safety system for mitigation.
The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition and safe storage and handling of fuel will be the only operations performed, and therefore, bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation will no longer be credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.
The proposed changes to delete and/or modify certain [requirements of the] TMI RFOL, TS, or CLB [Current Licensing Basis] have no impact on facility SSCs affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent irradiated fuel itself. The removal of TS that are related only to the operation of the nuclear reactor, or only to the prevention, diagnosis, or mitigation of reactor related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and TMI will no longer be authorized to operate the reactor.
The proposed modification or deletion of requirements of the TMI RFOL, TS, and CLB [does] not affect systems credited in the accident analysis for the remaining credible DBAs at TMI. The proposed RFOL and PDTS will continue to require proper control and monitoring of safety significant parameters and activities. The TS regarding SFP water level and spent fuel storage is retained to preserve the current requirements for safe storage of irradiated fuel.
The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding, spent fuel racks, SFP integrity, and SFP water level). Since extended operation in a defueled condition and safe fuel handling will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.
The proposed changes involve deleting and/or modifying certain [requirements of the] RFOL, TS, and CLB once the TMI facility has been permanently shutdown and defueled. Because the 10 CFR part 50 license for TMI [will] no longer [authorize] operation of the reactor, or emplacement or retention of fuel into the reactor vessel with the certifications required by 10 CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The remaining postulated DBA events that could potentially occur at a permanently defueled facility would be a FHA, WGTR, and FCDA. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses.
The proposed changes are limited to those portions of the RFOL, TS, and CLB that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the RFOL, TS, and CLB are not credited in the existing accident analysis for the remaining applicable postulated accidents; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated design basis accidents involving the reactor will no longer be possible because the reactor will be permanently shutdown and defueled and TMI will no longer be authorized to operate the reactor.
Date of amendment request: September 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18271A009.
Description of amendment request: The amendment would modify the applicability for Technical Specification (TS) Section 3.3.6.2, “Secondary Containment Isolation Instrumentation,” Functions 3 and 4, related to reactor building and refueling floor ventilation exhaust, respectively. This change would be implemented in the fall of 2019.
The requested changes to TS Section 3.3.6.2 to revise the applicability of Functions 3 and 4 as proposed does not eliminate the design function associated with the radiation monitoring instrumentation. The Secondary Containment Isolation Instrumentation will continue to automatically initiate closure of appropriate Secondary Containment Isolation Valves (SCIVs) and start the Standby Gas Treatment (SGT) system as designed to limit fission product release during any postulated Design Basis Accidents (DBAs). These systems are not accident initiators. The proposed changes will continue to assure that these systems perform their design functions, which includes mitigating accidents. The proposed changes do not alter the physical design of any plant Structure, System, or Components (SSC); therefore, the proposed changes have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to Start Printed Page 58613DBAs does not change and remains as analyzed in the Updated Final Safety Analysis Report (UFSAR).
The requested changes to TS Section 3.3.6.2 to revise the applicability of Functions 3 and 4 as proposed does not adversely affect the design function associated with the radiation monitoring instrumentation. The proposed changes do not change any system operations or maintenance activities that would create the possibility of a new or different kind of accident from one previously evaluated. The Secondary Containment Isolation Instrumentation and SGT system will continue to function as designed. The proposed changes will continue to assure that these systems perform their design functions, which includes mitigating accidents. The proposed changes do not create new failure modes or mechanisms and no new accident precursors are created. The proposed changes do not alter the plant configuration (no new or different type of equipment is being installed) or require any new or unusual Operator actions. The proposed changes do not alter the safety limits or safety analysis assumptions associated with the operation of the plant. The proposed changes do not introduce any new failure modes or mechanisms that could result in a new accident. The proposed changes do not reduce or adversely affect the capabilities of any plant SSC in the performance of their safety function. Also, the response of the plant and the Operators following any DBA is unaffected by the proposed changes.
The requested changes to TS Section 3.3.6.2 to revise the applicability of Functions 3 and 4 as proposed does not alter the design capability associated with the radiation monitoring instrumentation. The proposed changes have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to DBAs does not change. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There is no change being made to safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes.
Date of amendment request: September 20, 2018. A publicly-available version is in ADAMS under Accession No. ML18263A199.
Description of amendment request: The amendment would make administrative changes to Technical Specification 4.4.2.1, “Inservice Tendon Surveillance Requirements.” The amendment would add the words “except where an alternative, exemption, or relief has been authorized by the NRC” to allow NRC-approved exceptions to the 10 CFR 50.55a requirements. Also, the amendment would add a note to exempt from the requirements of Surveillance Requirement 4.0.1.
The addition of the words “except where an alternative, exemption, or relief has been authorized by the NRC” to Technical Specification (TS) 4.4.2.1 (“lnservice Tendon Surveillance Requirements”) and the addition of the wording “The surveillance interval extension allowed per Surveillance Requirement 4.0.1 is not permitted” are administrative changes that have no impact on the accidents analyzed and are not an accident initiator. Since the changes do not impact any conditions that would initiate an accident, the probability or consequences of previously analyzed events is not increased.
The proposed changes do not involve the modification of any plant equipment or affect plant operation. The proposed changes will have no impact on any safety-related structures, systems, or components.
No safety-related equipment, safety function, or plant operation will be altered as a result of these proposed administrative changes. No new operator actions are created as a result of the proposed changes. These administrative changes have no impact on the accidents analyzed in the Updated Final Safety Analysis Report (UFSAR) and are not accident initiators. These proposed changes do not impact the U.S. Nuclear Regulatory Commission Staff's authority to review and grant exceptions. The addition of the wording “The surveillance interval extension allowed per Surveillance Requirement 4.0.1 is not permitted” has been added to address the concerns identified in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report [(Reference 3 of the licensee's letter dated September 20, 2018)].
Since these proposed changes do not impact any conditions that would initiate an accident, there is no possibility of a new or different kind of accident resulting from these changes. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed administrative changes do not affect any margins of safety. The margins of safety presently provided by the Technical Specifications remain unchanged. The proposed amendment does not affect the design of the facility or system operating parameters, does not physically alter safety-related systems, structures, or components (SSCs) and does not affect the method in which safety-related systems perform their functions.
Date of amendment request: September 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18275A323.
Description of amendment request: The proposed amendment would revise Start Printed Page 58614the Renewed Facility License and the Permanently Defueled Technical Specifications (PDTS) for FCS to reflect the requirements after removal of all remaining spent nuclear fuel from the spent fuel pool (SFP) and its transfer to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI).
The proposed amendment would modify the FCS renewed facility operating license and PDTS by deleting the portions of the license and PDTS that are no longer applicable to a facility with no spent nuclear fuel stored in the spent fuel pool, while modifying the remaining portions to correspond to all nuclear fuel stored within an ISFSI. This amendment becomes effective upon removal of all spent nuclear fuel from the FCS SFP and its transfer to dry cask storage within an ISFSI. The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:
3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or § 100.11 .
The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI, none of the SSCs at FCS are required to be relied on for accident mitigation. Therefore, none of the SSCs at FCS meet the definition of a safety-related SSCs stated in 10 CFR 50.2. The proposed deletion of requirements in the FCS PDTS does not affect systems credited in any accident analysis at FCS.
Chapter 14 of the FCS Defueled Safety Analysis Report (DSAR) described the design basis accident related to the SFP. These postulated accidents are predicated on spent fuel being stored in the SFP. With the removal of the spent fuel from the SFP, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Shift Manager, Certified Fuel Handler, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident associated with nuclear fuel. The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences. The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the SFP, and relocate certain administrative controls to the Quality Assurance Topical Report which is a licensee-controlled document. After the removal of the spent fuel from the SFP and transfer to the ISFSI, there are no spent fuel assemblies that remain in the SFP. Coupled with a prohibition against storage of fuel in the SFP, the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes.
The removal of all spent nuclear fuel from the SFP into storage in casks within an ISFSI, coupled with a prohibition against future storage of fuel within the SFP, removes the potential for fuel related accidents.
The design basis and accident assumptions within the FCS DSAR and the PDTS relating to safe management and safety of spent fuel in the SFP are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities.
The requirements for SSCs that have been deleted from the FCS PDTS are not credited in the existing accident analysis for any applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis.
Date of amendment request: September 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18270A360.
Description of amendment request: The proposed amendment would correct a non-conservative Technical Specification (TS) 3/4.8.2, “DC [Direct Current] Sources -Operating,” by revising the inter-cell resistance value listed in Surveillance Requirements (SRs) 4.8.2.1.b.2 and 4.8.2.1.c.3.
Performing the proposed changes in battery parameter surveillance testing and verification is not a precursor of any accident previously evaluated. Furthermore, these changes will help to ensure that the voltage and capacity of the batteries is such that they will provide the power assumed in calculations of design basis accident mitigation. Therefore, SCE&G concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes to the VCSNS TS SR do not involve any physical modification of the plant or how the plant is operated. No new or different type of equipment will be installed. The proposed changes involve surveillance testing and verification activities. No new failure modes/effects which could lead to an accident whose consequences exceed the consequences of accidents previously analyzed will be introduced by the changes to the TS SR.
Margin of safety is related to the confidence in the ability of the fission Start Printed Page 58615product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of the fuel cladding, reactor coolant, and containment systems will not be impacted by the proposed changes.
The proposed VCSNS revisions of the SRs ensure the continued availability and operability of the batteries. As such, sufficient DC capacity to support operation of mitigation equipment remains within the design basis. Therefore, SCE&G concludes that the proposed changes do not involve a significant reduction in the margin of safety.
Date of amendment request: October 8, 2018. A publicly-available version is in ADAMS under Accession No. ML18281A014.
Description of amendment request: The proposed amendment would revise the Surveillance Requirement (SR) of Technical Specification (TS) 4.4.6.2.2 (a) to allow the reactor coolant system (RCS) pressure isolation valve (PIV) leakage test to be extended to a performance-based frequency not to exceed 3 refueling outages (RFOs) or 60 months following two consecutive satisfactory tests.
The proposed change involves revising the VCSNS Unit 1, TS wording to reflect a performance-based surveillance testing interval for leakage testing of the RCS PIVs. Specifically, the proposed change revises TS surveillance requirement (SR) 4.4.6.2.2.a to test the RCS PIVs at a frequency from each RFO to a maximum of every third RFO or 60 months by verifying that each of the PIVs tested in the associated RFO based on performance are within the TS allowable leakage limits. The RCS PIVs are defined as two normally closed valves in series with the reactor coolant pressure boundary (RCPB), which separate the high-pressure RCS from an attached lower pressure system. Excessive PIV leakage could lead to overpressure of the low-pressure piping or components, potentially resulting in a LOCA [loss-of-coolant accident] outside of containment.
TS SR 4.4.6.2.2.a for RCS PIVs provides added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent ISLOCA [intersystem loss-of-coolant accident]. The RCS PIV allowable leakage limit applies to each individual valve. This proposed change does not revise any of the TS RCS PIV allowable leakage limits. In addition, the RCS PIVs will continue to be tested per the VCSNS Inservice Testing Program in accordance with Title 10, Code of Federal Regulations (CFR), Section 50.55a, “Codes and standards.” The activity does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. By transitioning to a performance-based leakage testing interval, these valves will continue to be demonstrated operationally ready and reliable. In the event of a PIV leakage test failure, PIV testing would require the component to return to the initial interval of every RFO until good performance is re-established. Therefore, there is no impact on the assurance that the RCS PIVs will be able to perform their safety function(s).
The proposed change involves revising the VCSNS TS wording to reflect a performance-based surveillance testing interval for leakage testing of the RCS PIVs from each RFO to a maximum of every third RFO or 60 months based on valve performance. The technical testing methodology and associated acceptance criteria remain unchanged. The change in the testing frequency is a performance-based approach, which has been demonstrated acceptable in numerous applications across the industry (RCS PIV testing, 10 CFR 50, Appendix J, Option B).
The testing requirements involved to periodically demonstrate the integrity of the RCS PIVs exist to ensure the plant's ability to mitigate the consequences of an accident. There are not any accident initiators or precursors affected by this change. The proposed TS change does not involve a physical change to the plant or the manner in which the plant is operated or controlled.
The proposed change involves revising the TS SR 4.4.6.2.2.a and associated TS Bases to reflect a performance-based surveillance testing frequency of the RCS PIVs from each RFO to a maximum of every third RFO or 60 months. The technical testing methodology and associated TS allowable leakage limits/acceptance criteria remain unchanged. The testing frequency uses a performance based approach, which has been demonstrated acceptable in numerous applications across the industry (RCS PIV testing, 10 CFR 50, Appendix J, Option B). Thus, this amendment request does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The RCS PIVs will continue to be tested per the VCSNS Inservice Testing Program in accordance with 10 CFR 50.55a.
The primary reason for performance-based PIV test intervals is to eliminate unnecessary thermal cycles. The VCSNS program for monitoring fatigue due to operational cycles and transients consists of review, evaluation, and documentation of RCS operational transients/cycles based on recorded plant operating parameters (i.e., temperature, pressure, flow) for compliance with Technical Specification Sections 3.5.2, 3.5.3, and 5.7.1.
An additional reason for requesting performance-based PIV test intervals is dose reduction to conform with NRC and industry As Low As Reasonably Achievable (ALARA) radiation dose principles. The nominal fuel cycle lengths at VCSNS, Unit 1, are 18 months. However, since RFOs may be scheduled slightly beyond 18 months, a 60-month period is used to provide a bounding timeframe to encompass three RFOs. The review of recent historical data identified that PIV testing each RFO results in a total personnel dose of approximately 300 millirem (milli-Roentgen Equivalent Man, or mrem). Assuming all of the PIVs remain classified as good performers, the proposed extended test intervals would provide for a savings of approximately 600 mrem over an approximate 60-month period (three RFOs).
The proposed surveillance interval extension for the RCS PIVs is based on the performance of the PIVs. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The design, operation, testing methods, and acceptance criteria for the RCS PIV testing specified in applicable codes and standards will continue to be met.
Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius LLP, 1111 Pennsylvania Avenue NW, Washington, DC 20004.Start Printed Page 58616
Date of amendment request: October 19, 2018. A publicly-available version is in ADAMS under Accession No. ML18292A660.
Description of amendment request: The requested amendment proposes to depart from certified AP1000 Design Control Document (DCD) Tier 2* material that has been incorporated into the Updated Final Safety Analysis Report (UFSAR). Specifically, the proposed departure consists of changes to Tier 2* information in the UFSAR (which includes the plant-specific DCD information) to change the vertical reinforcement information provided in the VEGP Unit 3 column line 1 wall from elevation 135′-3″ to 137′-0″.
As described in UFSAR Subsection 3H.5.1.1, the exterior wall at column line 1 (Wall 1) is located at the south end of the auxiliary building. It is a reinforced concrete wall extending from the basemat at elevation 66′-6″ to the roof at elevation 180′-0″. Deviations were identified in the constructed wall from the design requirements. The proposed change modifies the vertical reinforcement information provided in the VEGP Unit 3 Wall 1 from elevation 135′-3″ to 137′- 0″. This change maintains conformance to the [American Concrete Institute (ACI)] 318-11 and ACI 349-01 codes and has no adverse impact on the seismic response of Wall 1. Wall 1 continues to withstand the design basis loads without loss of structural integrity or the safety-related functions. The proposed change does not affect the operation of any system or equipment that initiates an analyzed accident or alter any SSC [structures, systems, and components] accident initiator or initiating sequence of events.
This change does not adversely affect the design function of the VEGP Unit 3 Wall 1 or the SSCs contained within the auxiliary building. This change does not involve any accident initiating components or events, thus leaving the probabilities of an accident unaltered.
The proposed change modifies the vertical reinforcement information provided in the VEGP Unit 3 Wall 1 from elevation 135′-3″ to 137′-0″. As demonstrated by the continued conformance to the applicable codes and standards governing the design of the structures, the wall withstands the same effects as previously evaluated. The proposed change does not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed change does not adversely affect the design function of the auxiliary building Wall 1 or any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or non-safety-related equipment. This change does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.
The proposed change modifies the vertical reinforcement information provided in the VEGP Unit 3 Wall 1 from elevation 135′-3″ to 137′-0″. This change maintains conformance to the ACI 318-11 and ACI 349-01 codes. The change to the vertical reinforcement elevation 135′-3″ to 137′-0″ does not change the performance of the affected portion of the auxiliary building for postulated loads. The criteria and requirements of ACI 349-01 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to criteria and requirements in ACI 349-01 and therefore, maintains the margin of safety. The change does not alter any design function, design analysis, or safety analysis input or result, and sufficient margin exists to justify departure from the Tier 2* requirements for the wall. As such, because the system continues to respond to design basis accidents in the same manner as before without any changes to the expected response of the structure, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes. Accordingly, no significant safety margin is reduced by the change.
Date of amendment request: October 11, 2018. A publicly-available version is in ADAMS under Accession Nos. ML18284A447.
Description of amendment request: The requested amendment proposes changes to plant-specific Design Control Document (DCD) Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) that involve changes to combined license (COL) Appendix C, and corresponding changes to plant-specific Tier 1 information. The changes would revise the COL to relocate the power operated relief valves in the COL Appendix C, Inspections, Tests, Analyses, and Acceptance Criteria and in the UFSAR. An initial Federal Register notice was published on September 19, 2018 (83 FR 47375), providing an opportunity to comment, request a hearing, and petition for leave to intervene for a License Amendment Request (LAR) for the VEGP COLs. The licensee has submitted a revision, dated October 11, 2018, to the original LAR that was dated August 10, 2018. This revision increases the scope of the original LAR. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, Appendix D, design certification rule is also requested for the plant-specific DCD Tier 1 departures.
The proposed changes do not affect the operation or reliability of any system, structure or component (SSC) required to maintain a normal power operating condition or to mitigate anticipated transients without safety-related systems. With the proposed changes, the PORV [Power Operated Relief Valve] block valves are still able to perform the safety-related functions of containment isolation, steam generator isolation, and steam generator relief isolation. There is no Start Printed Page 58617change to the PORV block valves safety class or safety-related functions.
The relocation of the branch line in which the PORV block valves are installed in allows the PORV block valves to be closer to the containment penetration and maintain compliance with General Design Criterion (GDC) 57 for locating containment isolation valves as close to the containment as practical.
There is no impact to Chapter 15 evaluations. Changes to the PORV block valve and line size do not impact the mass releases to the atmosphere during a Steam Generator Tube Rupture accident. The mass release is limited by the PORV which is more restrictive than the PORV block valve and line size.
There is no impact to any assumed leakage through the PORV line. The existing 12-inch PORV has a design function to limit leakage through the PORV line. Increasing the PORV block valve to 12 inches will increase the leakage through the PORV block valve however it will be that same leakage rate as the 12-inch PORV. Therefore, the leakage rate through the PORV line does not increase and there is no impact to radiation doses.
There is no impact to the assumptions or analysis in the completed safety analysis for radiation doses as a result of the change.
There is no impact to the conclusions of the Pipe Rupture Hazard Analysis (PRHA) because the PORV line is Break Exclusion Zone (BEZ) piping. The proposed changes do not result in any new postulated break locations. Updated analyses confirm that the integrity of the wall adjacent to the MCR [main control room] is unaffected by a postulated main steam line break that causes the PORV line to impact the wall.
There is no change to the valve motor operator. The current motor operator is sufficient to operate the new 12-inch globe valve. Therefore, there is no impact to the Class 1E dc [direct current] and UPS [uninterruptable power supply] System (IDS) battery sizing. There is no change to the valve stroke time, therefore there is no impact to valve open/closure times.
The proposed changes do not affect the operation of systems or equipment that could initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. With the proposed changes, the PORV block valves are still able to perform the safety related functions of containment isolation, steam generator isolation, and steam generator relief isolation. There is no change to the PORV block valves safety class or safety-related functions.
There is no impact to the conclusions of the Pipe Rupture Hazard Analysis (PRHA) because the PORV line is Break Exclusion Zone (BEZ) piping. The proposed changes do not result in any new postulated break locations. Updated analyses confirm that the integrity of the wall adjacent to the MCR is unaffected by a postulated main steam line break that causes the PORV line to impact the wall.
There is no change to the valve motor operator. The current motor operator is sufficient to operate the new 12-inch globe valve. Therefore, there is no impact to the Class 1E dc and UPS System (IDS) battery sizing. There is no change to the valve stroke time, therefore there is no impact to valve open/closure times.
The proposed changes do not affect existing safety margins. With the proposed changes, the PORV block valves are still able to perform the safety-related functions of containment isolation, steam generator isolation, and steam generator relief isolation. There is no change to the PORV block valves safety class or safety-related functions.
The piping analysis for the affected piping has been revised in accordance with the requirements of the UFSAR. All stresses and interface loads remain acceptable and within the limits described in the UFSAR. The piping support calculations have been revised using the load combinations prescribed in the UFSAR, and the critical interaction ratio for each support is less than 1.0; therefore, a positive design margin exists. The proposed changes did not affect any of the piping packages chosen (as listed in the UFSAR) to demonstrate piping design for piping design acceptance criteria closure. There is no impact to the conclusions of the Pipe Rupture Hazard Analysis (PRHA) because the PORV line is Break Exclusion Zone (BEZ) piping. The proposed changes do not result in any new postulated break locations. Updated analyses confirm that the integrity of the wall adjacent to the MCR is unaffected by a postulated main steam line break that causes the PORV line to impact the wall. The piping and components downstream of the PORV are nonsafety-related and are not affected by this activity.
The structural concrete floors and walls which make up the bounds of the affected rooms were analyzed for the downstream impacts due to the proposed changes. The results conclude that the applicable acceptance criteria of the UFSAR are met. All applicable load combinations shown in the UFSAR were considered. Critical sections defined in the UFSAR within the scope of analysis remain unchanged along with the typical reinforcement configuration presented in the UFSAR. Therefore, all structural evaluations are within the bounds of the acceptance criteria and meet the licensing requirements imposed in the UFSAR.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.Start Printed Page 58618
Date of amendment request: May 14, 2018. A publicly available version is in ADAMS under Accession No. ML18138A232.
Description of amendment request: The proposed amendment would modify the WBN, Unit 2, Technical Specification (TS) 5.7.2.12, “Steam Generator (SG) Program,” and TS 5.9.9, “Steam Generator Tube Inspection Report,” to use the voltage-based alternate repair criteria (ARC) specified in the guidelines contained in Generic Letter (GL) 95-05, “Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking.”
Allowing the use of alternate repair criteria as proposed in this amendment request does not involve a significant increase in the probability or consequence of an accident previously evaluated.
Tube burst criteria are inherently satisfied during normal operating conditions due to the proximity of the TSP [tube support plates]. Test data indicates that tube burst cannot occur within the TSP, even for tubes, which have 100% through-wall electric discharge machining (EDM) notches, 0.75 inches long, provided that the TSP is adjacent to the notched area. Because tube-to-tube support plate proximity precludes tube burst during normal operating conditions, use of the criteria must retain tube integrity characteristics, which maintain a margin of safety of 1.4 times the bounding faulted condition [i.e., main steam line break (MSLB)] differential pressure of 2405 psig. GL 95-05 recommends that maintenance of a safety factor of 1.4 times the MSLB pressure differential, consistent with the structural limits in Regulatory Guide (RG) 1.121, on tube burst is satisfied by 3/4-inch diameter tubing with bobbin coil indications with signal amplitudes less than the tube structural limit (VSL) of 6.03 volts, regardless of the indicated depth measurement. At the FDB [flow distribution baffles], a safety factor of three against the normal operating condition ΔP is applied. A voltage of VSL = 3.81 volts satisfies the burst capability recommendation at the FDB.
The upper voltage repair limit (VURL) will be determined prior to each outage using the most recently approved NRC database to determine the VSL. The structural limit is reduced by allowances for nondestructive examination (NDE) uncertainty (VNDE) and growth (VG) to establish VURL.
Relative to the expected leakage during accident condition loadings, it has been previously established that a postulated MSLB outside of containment but upstream of the main steam isolation valves (MSIVs) represents the most limiting radiological condition relative to the alternate voltage-based repair criteria. In support of implementation of the revised repair limit, TVA will determine whether the distribution of cracking indications at the tube support plate intersections during future cycles are projected to be such that primary to secondary leakage would result in site boundary doses within a fraction of the 10 CFR 100 guidelines or control room doses within the 10 CFR 50, Appendix A, General Design Criterion (GDC) 19 limit. A separate calculation has determined this allowable MSLB leakage limit to be four gallons per minute (gpm) in the faulted loop.
The methods for calculating the radiological dose consequences for this postulated MSLB are consistent with the WBN dual-unit Updated Final Safety Analysis Report (UFSAR) Chapter 15.
In summary, the calculated radiological consequences in the control room and at the exclusion area boundary and the low population zone are in compliance with the guidelines in the Standard Review Plan, Chapter 15, and the regulations in 10 CFR 50, Appendix A, GDC 19, and 10 CFR 100 reported for the postulated steamline break event. Therefore, it is concluded that the proposed changes do not result in a significant increase in the radiological consequences of an accident previously analyzed.
Consistent with the guidance of GL 95-05, Section 2.c, the WBN Unit 2 MSLB leak rate analysis would be performed, prior to returning the SGs to service, based on either the projected next end-of-cycle (EOC) voltage distribution or the actual measured bobbin voltage distribution. The method to be used for the first outage when ODSCC [outside diameter stress corrosion cracking] indication growth rates are available will be based on the indications found during that outage. As noted in GL 95-05, it may not always be practical to complete EOC calculations prior to returning the SGs to service. Under these circumstances, it is acceptable to use the actual measured bobbin voltage distribution instead of the projected EOC voltage distribution to determine whether the reporting criteria are being satisfied.
Therefore, the voltage-based ARC at WBN Unit 2 does not adversely affect SG tube integrity and implementation is shown to result in acceptable radiological dose consequences. Therefore, the proposed TS change does not result in a significant increase in the probability or consequences of an accident previously evaluated within the WBN Unit 2 UFSAR.
Implementation of the proposed SG tube voltage-based ARC does not introduce any changes to the plant design basis. Neither a single nor multiple tube rupture event would be expected in an SG in which the repair limit has been applied (during all plant conditions).
The bobbin probe voltage-based tube repair criteria of 1.0 volt is supplemented by: enhanced eddy current inspection guidelines to provide consistency in voltage normalization, a 100 percent eddy current inspection sample size at the tube support plate elevations, and rotating probe coil (RPC) or equivalent inspection requirements for the larger indications left in service to characterize the principal degradation as ODSCC.
As SG tube integrity upon implementation of the 1.0 volt repair limit continues to be maintained through in-service inspection and primary to secondary leakage monitoring, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
The use of the voltage-based bobbin probe tube support plate elevation repair criteria at WBN Unit 2 maintains SG tube integrity commensurate with the guidance of RG 1.121. RG 1.121 describes a method acceptable to the NRC for meeting GDCs 14, 15, and 32 by reducing the probability or the consequences of SG tube rupture. This reduction is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by in-service inspection, for which tubes with unacceptable cracking should be removed from service. Upon implementation of the proposed criteria, even under the worst-case conditions, the occurrence of ODSCC at the TSP elevations is not expected to lead to an SG tube rupture event during normal or faulted plant conditions. The EOC distribution of crack indications at the tube support plate elevations is confirmed to result in acceptable primary to secondary leakage during all plant conditions and that radiological consequences are not adversely impacted.
Implementation of the TSP intersection voltage-based repair criteria will decrease the number of tubes that must be plugged. The installation of SG tube plugs reduces the reactor coolant system flow margin. Thus, implementation of the 1.0 volt repair limit will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.
Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.Start Printed Page 58619
Date of amendment request: February 28, 2018. A publicly-available version is in ADAMS under Accession No. ML18060A337.
Description of amendment request: The proposed amendments would modify the WBN, Units 1 and 2, Technical Specification (TS) 3.8.9, to add a new Condition C with an 8-hour completion for performing maintenance on the opposite unit's vital bus when the opposite unit is in Mode 5, Mode 6, or defueled. The proposed change would allow greater operational flexibility for two-unit operation at WBN.
The proposed change modifies the Required Actions for the opposite unit's 120-volt (V) alternating current (AC) vital bus system. This change will not affect the probability of an accident, because the distribution system is not an initiator of any accident sequence analyzed in the UFSAR [updated final safety analysis report]. Rather, the opposite unit's distribution system support equipment is used to mitigate accidents. The consequences of an analyzed accident will not be significantly increased because the minimum requirements for distribution systems will be maintained to ensure the availability of the required power to mitigate accidents assumed in the UFSAR. Operation in accordance with the proposed TS will ensure that sufficient onsite electrical distribution systems are operable as required to support the unit's required features. Therefore, the mitigating functions supported by the onsite electrical distribution systems will continue to provide the protection assumed by the accident analysis. The integrity of fission product barriers, plant configuration, and operating procedures as described in the UFSAR will not be affected by the proposed changes. Thus, the consequences of previously analyzed accidents will not increase by implementing these changes.
The proposed change modifies the Required Actions for the opposite unit's 120V AC vital bus system. This change will not physically alter the plant (no new or different type of equipment will be installed). The proposed change will maintain the minimum requirements for onsite electrical distribution systems to ensure the availability of the equipment required to mitigate accidents assumed in the UFSAR.
The proposed change modifies the Required Actions for the opposite unit's 120V AC vital bus system. The margin of safety is not affected by this change because the minimum requirements for onsite electrical distribution systems will be maintained to ensure the availability of the required power to shutdown the reactor and maintain it in a safe shutdown condition after an AOO [anticipated operational occurrence] or a postulated DBA [design-basis accident].
Date of amendment request: June 28, 2017, as supplemented by letters dated July 20 and September 14, 2017; and January 18, February 16, and April 13, 2018.
Brief description of amendment: The amendment revised the Technical Specifications (TSs) for fuel storage criticality to account for the use of neutron absorbing spent fuel pool rack inserts and soluble boron for the purpose of criticality control in the boiling-water reactor storage racks that currently credit Boraflex.
Amendment No.: 167. A publicly-available version is in ADAMS under Accession No. ML18204A286; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Date of initial notice in Federal Register: December 5, 2017 (82 FR 57481). The supplemental letters dated July 20 and September 14, 2017; and January 18, February 16, and April 13, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no Start Printed Page 58620significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 22, 2018.
Date of amendment request: October 23, 2017, as supplemented by letters dated November 15, 2017, and June 27, 2018.
Date of issuance: October 30, 2018.
Effective date: As of its date of issuance and shall be implemented at the beginning of the next refueling outage scheduled for May 2019.
Amendment No.: 251. A publicly-available version is in ADAMS under Accession No. ML18255A350; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: January 16, 2018 (83 FR 2227). The supplemental letter dated June 27, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 30, 2018.
Date of amendment request: October 2, 2017, as supplemented by letters dated April 26 and August 10, 2018.
Brief description of amendment: The amendment revised the ANO-1 Technical Specification (TS) Bases for TS 3.7.5, “Emergency Feedwater (EFW) System,” to identify the conditions in which TS 3.7.5, Condition A, 7-day Completion Time (CT) and Condition C, 24-hour CT should apply to the ANO-1 turbine-driven EFW pump steam supply valves.
Date of issuance: October 24, 2018.
Amendment No.: 261. A publicly-available version is in ADAMS under Accession No. ML18260A339; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: The amendment revised the TS Bases.
Date of initial notice in Federal Register: December 5, 2017 (82 FR 57473). The supplemental letters dated April 26 and August 10, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 24, 2018.
Date of amendment request: November 16, 2017, as supplemented by letter dated March 29, 2018.
Brief description of amendment: The amendment revised the Oyster Creek Renewed Facility Operating License and the associated Technical Specifications (TS) to Permanently Defueled Technical Specifications consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.
Date of issuance: October 26, 2018.
Effective date: The license amendment is effective on November 16, 2018, and shall be implemented in 60 days from the effective date.
Amendment No.: 295. A publicly-available version is in ADAMS under Accession No. ML18227A338; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-16: The amendment revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: January 16, 2018 (83 FR 2229). The supplemental letter dated March 29, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 26, 2018.
Date of amendment request: June 25, 2018, as supplemented by letter dated August 29, 2018.
Brief description of amendment: The amendment revised the R. E. Ginna Nuclear Power Plant's Technical Specification (TS) 3.1.4, “Rod Group Alignment Limits”; TS 3.1.5, “Shutdown Bank Insertion Limit”; TS 3.1.6, “Control Bank Insertion Limits”; and TS 3.1.7, “Rod Position Indication,” consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-547, Revision 1, “Clarification of Rod Position Requirements,” dated March 4, 2016.
Date of issuance: October 31, 2018.
Amendment No.: 131. A publicly-available version is in ADAMS under Accession No. ML18295A630; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-18: The amendment revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: July 31, 2018 (83 FR 36976). The supplemental letter dated August 29, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's Start Printed Page 58621original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 31, 2018.
Date of amendment request: April 25, 2018.
Brief description of amendments: The amendments revised the Technical Specification (TS) requirements for inoperable snubbers for each facility. The amendments also made other administrative changes to the TS.
Date of issuance: October 29, 2018.
Amendment Nos.: Clinton—220 (Unit 1); Dresden—259 (Unit 2), 252 (Unit 3); LaSalle—231 (Unit 1), 217 (Unit 2); and Quad Cities—271 (Unit 1), 266 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18254A367. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, and DPR-30: The amendments revised the Facility Operating Licenses and TS.
Date of initial notice in Federal Register: June 19, 2018 (83 FR 28460).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 29, 2018.
Date of amendment request: August 30, 2017, as supplemented by letters dated October 24, 2017; and May 7, June 6, August 10, and August 22, 2018.
Date of issuance: October 25, 2018.
Amendment Nos.: 321 (Unit 2) and 324 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML18263A232; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: November 21, 2017 (82 FR 55404). The supplemental letters dated May 7, June 6, August 10, and August 22, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 25, 2018.
Date of amendment request: February 25, 2016, as supplemented by letters dated April 3, 2017, and January 11, January 18, June 21, and August 27, 2018.
Brief description of amendments: The amendments revised the Calvert Cliffs Technical Specifications (TS) related to completion times for required actions to provide the option to calculate longer risk-informed completion times. The amendments also added a new program, the “Risk Informed Completion Time Program,” to TS Section 5.5, “Programs and Manuals.”
Effective date: As of the date of its issuance and shall be implemented within 180 days.
Amendment Nos.: 326 (Unit 1) and 304 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18270A130; documents related to these amendments are listed in the safety evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-53 and DPR-69: The amendments revised the Renewed Facility Operating Licenses and TS.
Date of initial notice in Federal Register: September 4, 2018 (83 FR 44920). The supplemental letter dated August 27, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained in a safety evaluation dated October 30, 2018.
Date of amendment request: August 2, 2018.
Brief description of amendments: The amendments revised the Technical Specifications (TS) by removing Figure 5.1-1, “Site Area Map”; removing Technical Specification references to Figure 5.1-1; and adding a site description.
Date of issuance: November 2, 2018.
Amendment Nos.: 246 (Unit No. 1) and 197 (Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML18274A224; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-67 and NPF-16: The amendments revised the Renewed Facility Operating Licenses and TS.
The Commission's related evaluation of the amendments is contained in a Start Printed Page 58622Safety Evaluation dated November 2, 2018.
Date of amendment request: November 10, 2017.
Brief description of amendment: The amendment revised the Technical Specifications (TS) for DAEC to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, “Revise Secondary Containment Surveillance Requirements,” dated November 10, 2017 (ADAMS Accession No. ML17318A240).
Amendment No.: 307. A publicly-available version is in ADAMS under Accession No. ML18241A383; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-49: The amendment revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: February 27, 2018 (83 FR 8517).
Date of amendment request: October 20, 2017, as supplemented by letters dated June 1 and September 11, 2018.
Brief description of amendment: The amendment revised the Monticello Technical Specification (TS) to adopt Technical Specification Task Force (TSTF) Traveler TSTF-542, “Reactor Pressure Vessel Water Inventory Control.”
Amendment No.: 198. A publicly-available version is in ADAMS under Accession No. ML18250A075; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-22. The amendment revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: December 19, 2017 (82 FR 60228). The supplemental letters dated June 1 and September 11, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 29, 2018.
Date of amendment request: September 21, 2017, as supplemented by letters dated June 27, July 19, and September 6, 2018.
Brief description of amendment: The amendment revised the Hope Creek Technical Specifications (TS) by replacing the existing specifications related to “operation with a potential for draining the reactor vessel” with revised requirements for reactor pressure vessel water inventory control to protect Safety Limit 2.1.4. Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel. The amendment adopted changes with variations, as noted in the license amendment request, and is based on the NRC-approved safety evaluation for Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, “Reactor Pressure Vessel Water Inventory Control,” dated December 20, 2016.
Effective date: As of the date of issuance and shall be implemented prior to entering Operating Condition 4 for the next Hope Creek refueling outage schedule for fall 2019 (H1R22).
Amendment No.: 213. A publicly-available version is in ADAMS under Accession No. ML18260A203; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License No. NPF-57: The amendment revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: January 30, 2018 (83 FR 4294). The supplemental letters dated June 27, July 19, and September 6, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
Date of amendment request: September 12, 2017, as supplemented by letter dated April 5, 2018.
Brief description of amendments: The amendments revised Technical Specification (TS) 5.5.17, “Containment Leakage Rate Testing Program,” for Vogtle to (1) increase the existing Type A integrated leakage rate test interval from 10 to 15 years; (2) extend the Type C containment isolation valve leaking testing to a 75-month frequency; (3) adopt the use of American National Standards Institute/American Nuclear Society 56.8-2002, “Containment System Leakage Testing Requirements”; and (4) adopt a more conservative grace interval for Type A, B, and C tests.
Amendment Nos.: 197 (Unit 1) and 180 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18263A039; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-68 and NPF-81: The amendments revised the Renewed Facility Operating Licenses and TS.
Date of initial notice in Federal Register: December 5, 2017 (82 FR 57474). The supplemental letter dated April 5, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
No significant hazards consideration comments received: No.Start Printed Page 58623
Date of amendment request: April 13, 2018, as supplemented by letter dated August 10, 2018.
Description of amendment: The amendment authorized changes to the VEGP Units 3 and 4 Combined Operating License (COL) Appendix A, Technical Specifications (TS). The amendment authorized departures from associated Updated Final Safety Analysis Report information (which includes the plant specific design control document Tier 2 information) with changes which conform with the authorized TS changes.
Date of issuance: October 11, 2018.
Amendment Nos.: 146 (Unit 3) and 145 (Unit 4). A publicly-available version is in ADAMS under Accession No. ML18248A137; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Combined Licenses Nos. NPF-91 and NPF-92: The amendment revised the Facility Combined Licenses and TS.
Date of initial notice in Federal Register: June 27, 2018 (83 FR 30199). The supplemental letter dated August 10, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated October 11, 2018.
Date of amendment request: October 11, 2017.
Brief description of amendment: The amendment revised Technical Specification (TS) 3.3.1, Table 3.3.1-1, “Reactor Trip System (RPS) Instrumentation,” to increase the values for the nominal trip setpoint and the allowable value for Function 14.a, “Turbine Trip − Low Fluid Oil Pressure.” The changes are due to the planned replacement and relocation of the pressure switches from the low pressure auto-stop trip fluid oil header to the high pressure turbine electrohydraulic control (EHC) oil header. The changes are needed due to the higher EHC system operating pressure.
Effective date: As of the date of issuance and shall be implemented no later than startup from the Unit 2 refueling outage scheduled for spring 2019.
Amendment No.: 22. A publicly-available version is in ADAMS under Accession No. ML18255A156; documents related to the amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-96: The amendment revised the Facility Operating License and TS.
Date of initial notice in Federal Register: March 13, 2018 (83 FR 10924).
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.Start Printed Page 58624
All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic Start Printed Page 58625storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.
Date of amendment request: September 5, 2018, as supplemented by letters dated September 20 and October 3, 2018.
Description of amendment: The amendments revised the CPNPP Technical Specification (TS) 3.8.4, “DC [Direct Current] Sources—Operating,” by adding a new REQUIRED ACTION to CONDITION B and an extended COMPLETION TIME on a one-time basis to repair two affected battery cells on the CPNPP Unit 1, Train B safety-related batteries.
Effective date: As of the date of issuance and shall be implemented immediately as of its date of issuance.
Amendment Nos.: Unit 1—170; Unit 2—170. A publicly-available version is in ADAMS under Accession No. ML18267A384; documents related to the amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Facility Operating Licenses and TS.
The license amendment request was originally noticed in the Federal Register on September 18, 2018 (83 FR 47203). Subsequently, by letters dated September 20 and October 3, 2018, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, on October 10, 2018 (83 FR 50971), the NRC published a second proposed NSHC determination, which superseded the original notice in its Start Printed Page 58626entirety. This included an individual 14-day notice for comments and provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by December 10, 2018, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated October 25, 2018.
Dated at Rockville, Maryland, this 8th day of November 2018.
[FR Doc. 2018-24894 Filed 11-19-18; 8:45 am]