2015-28347. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from October 10, 2015, to October 26, 2015. The last biweekly notice was published on October 27, 2015.

DATES:

Comments must be filed December 10, 2015. A request for a hearing must be filed by January 11, 2016.

ADDRESSES:

You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0253. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail].
• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT:

Lynn Ronewicz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: [phone], email: [e-mail].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2015-0253 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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0253.
• 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/​adams.html. To begin the search, select “ADAMS Public Documents” and then 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 [e-mail]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section of this document.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2015-0253, facility name, unit number(s), application date, and subject in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (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. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).

If a hearing is requested, and the Commission has not made a final determination 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 decide 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 held 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 any 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.

A State, local governmental body, federally-recognized Indian tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by December 28, 2015. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions for leave to intervene set forth in this section, except that under § 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).

If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by December 28, 2015.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, 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 participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). 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 storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [e-mail], or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by email to [e-mail], or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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 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, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document 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 delivery 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.

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​,, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: July 17, 2015. A publicly-available version is in ADAMS under Accession No. ML15232A017.

Description of amendment request: The proposed amendment corrects a usage problem with recently issued Amendment Nos. 382, 384, and 383 (ADAMS Accession No. ML13231A013), which precludes Oconee Nuclear Station Technical Specification (TS) 3.8.1, “AC [Alternating Current] Sources-Operating,” Condition H from being used as planned. The proposed change revises the note to TS 3.8.1 Required Actions L.1, L.2, and L.3, to remove the 12-hour time limitation when the second Keowee Hydroelectric Unit (KHU) is made inoperable for the purpose of restoring the KHU undergoing maintenance to OPERABLE status. Removal of the 12-hour time limitation allows use of the full 60-hour Completion Time of Required Action H.2 when the unit(s) have been in Condition C for greater than 72 hours and both units are made inoperable for purposes of restoring the KHU undergoing maintenance to OPERABLE status.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment revises the note to Technical Specification (TS) 3.8.1 Required Actions L.1, L.2, and L.3 to indicate the Required Actions are not required when the Condition is entered to restore a KHU to OPERABLE status. This change is consistent with Amendment Nos. 382, 384, and 383, which approved a cumulative 240 hours of allowed outage time over a 3-year period when both KHUs are inoperable when in the 45-day Completion Time of TS 3.8.1 Required Action C.2.2.5. The proposed TS change does not modify the reactor coolant system pressure boundary, nor make any physical changes to the facility design, material, or construction standards. The probability of any design basis accident (DBA) is not affected by this change, nor are the consequences of any DBA affected by this change. The proposed change does not involve changes to any structures, systems, or components (SSCs) that can alter the probability for initiating a LOCA [loss-of-coolant accident] event.

Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed TS change revises the note to TS 3.8.1 Required Actions L.1, L.2, and L.3 to indicate the Required Actions are not required when the Condition is entered to restore a KHU to OPERABLE status. Revision of the note allows the 60 hour Completion Time of TS 3.8.1 Condition H to limit the time that both KHUs are inoperable. The changes do not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. No new failure modes are identified, nor are any SSCs required to be operated outside the design bases.

Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed TS change revises the note to TS 3.8.1 Required Actions L.1, L.2, and L.3 to indicate the Required Actions are not required when the Condition is entered to restore a KHU to OPERABLE status. Revision of the note allows the 60 hour Completion Time of TS 3.8.1 Condition H to limit the time that both KHUs are inoperable. The proposed TS change does not involve: (1) A physical alteration of the Oconee Units; (2) the installation of new or different equipment; (3) operating any installed equipment in a new or different manner; (4) a change to any set points for parameters which initiate protective or mitigation action; or (5) any impact on the fission product barriers or safety limits.

Therefore, this request does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon Street—DEC45A, Charlotte, NC 28202-1802.

NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida

Date of amendment request: August 27, 2015. A publicly-available version is in ADAMS under Accession No. ML15246A231.

Description of amendment request: The amendment would approve changes to the Permanently Defueled Emergency Plan (PDEP) to reflect the planned use of an Independent Spent Fuel Storage Installation (ISFSI) located in the Crystal River Unit 3 Nuclear Plant Protected Area while the spent fuel pool contains spent fuel assemblies.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed site PDEP and PD EAL [Permanently Defueled Emergency Action Level] Bases Manual revisions are commensurate with the ongoing and anticipated reduction in radiological source term at the CR-3 site and reflects the addition of spent fuel being transferred to the ISFSI facility. These changes add the responsibility for responding to ISFSI emergencies to the CR-3 PDEP Shift Supervisor/Certified Fuel Handler, and accompanying changes to the PD EAL Bases Manual due to the creation of a potential or actual release path to the environment, degradation of one or more storage canisters or fuel assemblies due to environmental factors, and configuration changes that could cause challenges in removing the canister or fuel from storage.

There are no longer design basis accidents or postulated beyond design basis accidents that could result in doses to the public and the environment beyond the exclusion area boundary that would exceed the EPA PAGs [Protective Action Guidelines]. CR-3 was shut down on September 26, 2009, and will not be restarted. With the reactor permanently defueled, the spent fuel pool and its support systems are dedicated to spent fuel storage only. With the spent fuel in wet storage for some time, the spectrum of postulated accidents is much smaller than for an operational plant, with the majority of design basis accidents no longer possible. The only remaining credible design basis accident is the fuel handling accident, which does not result in exceeding the EPA Protective Action Guidelines at the exclusion area boundary. Spent fuel located in the spent fuel pools will be transferred to the ISFSI facility. Emergency Planning Zones beyond the exclusion area boundary and the associated protective actions are no longer required. No corporate personnel, personnel involved in off-site dose projections, or personnel with special qualifications are required to augment the ERO [Emergency Response Organization].

The credible events for the ISFSI facility remain unchanged. The indications of damage to a loaded Dry Shielded Canister CONFINEMENT BOUNDARY have been revised to be twice the design basis dose rate as described in Draft Amendment 14 to COC [Certificate of Compliance] 1004 Technical Specifications for the Standardized NUHOMS Horizontal Modular Storage System, Sections 5.2.4 `Radiation Protection Program' and 5.4.2 HSM [horizontal storage module] or HSM-H Dose Rate Evaluation Program (Reference 7), while in transit or HSM storage.

Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as indicated by the following on-contact radiation readings at some prescribed distance from the transfer cask or HSM:

1300 mrem/hr (gamma + neutron) on the radial surface of the fuel transfer cask while in transit to the ISFSI HSM

OR

1050 mrem/hr (gamma + neutron)—HSM Front Bird Screen

4 mrem/hr (gamma + neutron)—HSM Outside Door

40 mrem/hr (gamma + neutron)—HSM End Shield Wall Exterior while in HSM storage.

This change is consistent with industry practices previously approved by the NRC to distinguish whether a degraded containment barrier condition exists.

The probability of occurrence of previously evaluated accidents is not increased, since most previously analyzed accidents can no longer occur and the probability of the remaining credible design basis accident is unaffected by the proposed amendment.

The deletion of the Communicator position does not impact Emergency Notifications from the plant since the Emergency Coordinator has shown the capability to perform this function. This function is not involved in operations or evolutions that could cause an accident since it is not performed until after the emergency is declared, and has no effect on accident mitigation.

Therefore, the proposed changes do not affect any plant system, the operation and maintenance of CR-3 and the ISFSI facility, or increase 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 proposed changes have no impact on facility structures, systems, or components (SSCs) affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. Additionally, the proposed changes have no impact on a Fuel Handling Accident, which is the remaining credible design basis accident evaluated. The CR-3 PDEP is applicable for the plant's defueled condition. There is no impact on the prevention, diagnosis, or mitigation of reactor-related transients as there are no longer any reactor-related accidents. Accidents cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor is permanently shut down and defueled, and CR-3 is no longer authorized to operate the reactor.

There are no longer credible events that would result in doses to the public beyond the exclusion area boundary that would exceed the EPA [Environmental Protection Agency] PAGs. Spent fuel waste will be transferred to the ISFSI facility. Emergency Planning Zones beyond the site boundary and the associated protective actions are no longer required. No corporate personnel, personnel involved in offsite dose projections, or personnel with special qualifications are required to augment the ERO.

The credible events for the ISFSI facility remain unchanged. The indications of damage to a loaded Dry Shielded Canister CONFINEMENT BOUNDARY have been revised to be twice the design basis dose rate as described in Draft Amendment 14 to COC 1004 Technical Specifications for the Standardized NUHOMS Horizontal Modular Storage System, Sections 5.2.4 `Radiation Protection Program' and 5.4.2 HSM or HSM-H Dose Rate Evaluation Program (Reference 7), while in transit or HSM storage.

Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as indicated by the following on-contact radiation readings at some prescribed distance from the transfer cask or HSM:

1300 mrem/hr (gamma + neutron) on the radial surface of the fuel transfer cask while in transit to the ISFSI horizontal storage module (HSM)

OR

1050 mrem/hr (gamma + neutron)—HSM Front Bird Screen

4 mrem/hr (gamma + neutron)—HSM Outside Door

40 mrem/hr (gamma + neutron)—HSM End Shield Wall Exterior while in HSM storage.

This change is consistent with industry practices previously approved by the NRC to distinguish whether a degraded containment barrier condition exists. The proposed amendment does not introduce a new mode of plant operation or new accident pre-cursors, does not involve any physical alterations to plant configurations, or make changes to plant system set points that initiate a new or different kind of accident.

The deletion of the Communicator position does not impact Emergency Notifications from the plant since the Emergency Coordinator has shown the capability to perform this function. This function is not involved in operations or evolutions that could cause or create new or different kinds of accidents since the communication of Emergency Notifications is not performed until after the emergency is declared and cannot affect an accident or event already in progress.

Therefore, the proposed changes have no direct effect on any plant system, the operation and maintenance of CR-3 or the ISFSI facility, or create the possibility of a new or different kind of accident.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes have no direct effect on any plant system, do not involve any physical plant limit or parameter, License Condition, Technical Specification Limiting Condition of Operability or operating philosophy, and therefore cannot affect any margin of safety. The margin of safety is maintained by conforming to the CR-3 Technical Specifications or the ISFSI Technical Specifications. The proposed CR-3 PDEP and PD EAL Bases Manual revisions are commensurate with the on-going and anticipated reduction in radiological source term at the CR-3 site and reflect spent fuel being transferred to the ISFSI facility. These changes add the responsibility for implementing the emergency plan for the ISFSI facility to the Shift Supervisor/Certified Fuel Handler.

The only remaining credible accident for CR-3, while the SFP is operable and prior to the transference of all spent fuel to dry shielded canisters, is a fuel handling accident. The proposed amendment does not adversely affect the inputs or assumptions of any design basis analysis that impact the fuel handling accident. There are no longer credible events that would result in doses to the public beyond the exclusion area boundary that would exceed the EPA PAGs. Emergency Planning Zones beyond the exclusion area boundary and the associated protective actions are no longer required. No corporate personnel, personnel involved in offsite dose projections, or personnel with special qualifications are required to augment the ERO. The credible events for the ISFSI facility remain unchanged. The indications of damage to a loaded Dry Shielded Canister CONFINEMENT BOUNDARY have been revised to be twice the design basis dose rate as described in Draft Amendment 14 to COC 1004 Technical Specifications for the Standardized NUHOMS Horizontal Modular Storage System, Sections 5.2.4 `Radiation Protection Program' and 5.4.2 HSM or HSM-H Dose Rate Evaluation Program (Reference 7), while in transit or HSM storage.

Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as indicated by the following on-contact radiation readings at some prescribed distance from the transfer cask or HSM:

1300 mrem/hr (gamma + neutron) on the radial surface of the fuel transfer cask while in transit to the ISFSI HSM

OR

1050 mrem/hr (gamma + neutron)—HSM Front Bird Screen

4 mrem/hr (gamma + neutron)—HSM Outside Door

40 mrem/hr (gamma + neutron)—HSM End Shield Wall Exterior while in HSM storage.

This change is consistent with industry practices previously approved by the NRC to distinguish whether a degraded containment barrier condition exists. The proposed changes are limited to the CR-3 PDEP and PD EAL Bases Manual and do not impact the safe storage of irradiated fuel. The proposed revisions do not affect any requirements for SSCs credited in the remaining analyses of applicable postulated accidents, and as such, do not affect the margin of safety associated with these accident analyses.

The deletion of the Communicator position does not impact Emergency Notifications from the plant since the Emergency Coordinator has shown the capability to perform this function. This function is not involved in design basis analyses or operations that could cause any decrease in any previously analyzed safety margin.

Therefore, the proposed changes do not create the possibility of reduction in any safety margin.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lara S. Nichols, 550 South Tryon Street, Charlotte NC 28202.

NRC Branch Chief: Bruce A. Watson, CHP.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: September 8, 2015. A publicly-available version is in ADAMS under Accession No. ML15258A185.

Description of amendment request: The proposed amendment would replace the Technical Specification (TS) Figure 4.1-1, “Site and Exclusion Area Boundaries and Low Population Zone,” with a text description in TS 4.1, “Site Location.” In addition, a typographical error would be corrected from “LGHR” to “LHGR” [Linear Heat Generation Rate] in TS 1.1, “Definitions.”

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change removes a figure, replaces that figure with a text description of the site location and corrects a typographical error. An administrative change such as this is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident with the incorporation of this administrative change are not different than the consequences of the same accident without this change. As a result, the consequences of an accident previously evaluated are not affected by this change.

Based on the above, it is concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not modify the plant design, nor does the proposed change alter the operation of the plant or equipment involved in either routine plant operation or in the mitigation of design basis accidents. The proposed change is administrative only.

Based on the above, it is concluded that the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change consists of an administrative change to remove a figure, replace that figure with a text description of the site location, and correct a typographical error. The change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside of the design basis.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Michael T. Markley.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: July 24, 2015. A publicly-available version is in ADAMS under Accession No. ML15246A408.

Description of amendment request: The amendment would make editorial corrections to Technical Specification (TS) Section 1.4, “Frequency.” Example 1.4-1 would be revised to be consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-485, Revision 0, “Correct Example 1.4-1.” In addition, Example 1.4-5 and Example 1.4-6 would be revised to correct typographical errors.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which 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 proposed changes are editorial in nature and have no effect on accident scenarios previously evaluated. The proposed changes consist of editorial corrections to TS Section 1.4, “Frequency,” that would make the Duane Arnold Energy Center (DAEC) TS consistent with the Standard Technical Specifications for General Electric BWR/4 Plants (NUREG-1433). The proposed changes do not affect initiating events for accidents previously evaluated and do not affect or modify plant systems or procedures used to mitigate the progression or outcome of those accident scenarios.

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 proposed changes are editorial in nature consisting of editorial corrections to TS Section 1.4, “Frequency.” The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.

The proposed changes do not introduce any new accident precursors, nor do they impose any new or different requirements or eliminate any existing requirements. The proposed changes do not alter assumptions made in the safety analysis.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed changes are editorial in nature consisting of editorial corrections to TS Section 1.4, “Frequency.” No setpoints at which protective actions are initiated are altered by the proposed changes. The proposed changes do not alter the manner in which the safety limits are determined. These changes are consistent with plant design and do not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: David L. Pelton.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of amendment request: August 6, 2015. A publicly-available version is in ADAMS under Accession No. ML15246A410.

Description of amendment request: The proposed amendment would resolve a 10 CFR part 21 condition concerning a potential to momentarily violate Reactor Core Safety Limit 2.1.1.1 during Pressure Regulator Failure Maximum Demand (Open) transient.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which 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 proposed change to the reactor steam dome pressure from 785 psig to 685 psig in TS [Technical Specification] SLs [Safety Limits] 2.1.1.1 and 2.1.1.2 does not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. The proposed change is in accordance with an NRC approved critical power correlation methodology and as such maintains required safety margins. The proposed change does not adversely affect accident initiators or precursors nor does it alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained.

The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not require any physical change to any plant SSCs nor does it require any change in systems or plant operations. The proposed change is consistent with the safety analysis assumptions and resultant consequences.

Therefore, the proposed change does 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 proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change.

The proposed change does not introduce any new accident precursors, nor does it impose any new or different requirements or eliminate any existing requirements. The proposed change does not alter assumptions made in the safety analysis.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. Evaluation of the 10 CFR part 21 condition by General Electric determined that there was no decrease in the safety margin, the Minimum Critical Power Ratio improves during the transient, and therefore is not a threat to fuel cladding integrity.

The proposed change to Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2 is consistent with, and within the capabilities of the applicable NRC approved critical power correlation, and thus continues to ensure that valid critical power calculations are performed. No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined. This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. William Blair, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: David L. Pelton.

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: June 12, 2015, as supplemented by letters dated August 11, 2015, and August 28, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15166A042, ML15223B277, and ML15240A017, respectively.

Description of amendment request: The amendments would revise the Point Beach Emergency Plan, to increase the staff augmentation times for Emergency Response Organization (ERO) response functions, from 30 and 60 minutes, to 60 minutes and 90 minutes, respectively. Additional changes include relocation of the Emergency Director and Emergency Action Level Monitor positions, and the addition of an Assistant Emergency Operations Facility Manager position.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which 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 proposed increase in staff augmentation times has no effect on normal plant operation or on any accident initiator or precursors and does not impact the function of plant structures, systems, or components (SCCs). The proposed change does not alter or prevent the ability of the ERO to perform their intended functions to mitigate the consequences of an accident or event. The ability of the ERO to respond adequately to radiological emergencies has been demonstrated as acceptable through a staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9.

Therefore, the proposed Emergency Plan 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 proposed change does not impact the accident analysis. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions. The proposed change does not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. This proposed change increases the staff augmentation response times in the Emergency Plan, which are demonstrated as acceptable through a staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9. The proposed change does not alter or prevent the ability of the ERO to perform their intended functions to mitigate the consequences of an accident or event.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed change is associated with the Emergency Plan staffing and does not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed change does not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed change. Safety analysis acceptance criteria are not affected by this proposed change. The revised Emergency Plan will continue to provide the necessary response staff with the proposed change. A staffing analysis and a functional analysis were performed for the proposed change on the timeliness of performing major tasks for the functional areas of Emergency Plan. The analysis concluded that an extension in staff augmentation times would not significantly affect the ability to perform the required Emergency Plan tasks. Therefore, the proposed change is determined to not adversely affect the ability to meet 10 CFR 50.54(q)(2), the requirements of 10 CFR 50 Appendix E, and the emergency planning standards as described in 10 CFR 50.47(b).

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: William Blair, Managing Attorney—Nuclear, Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, Juno Beach, FL 33408-0420.

NRC Branch Chief: David L. Pelton.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo County, California

Date of amendment request: September 16, 2015. A publicly-available version is in ADAMS under Accession No. ML15259A576.

Description of amendment request: The amendment would revise the Reactor Coolant System (RCS) minimum flow specified in Technical Specification (TS) 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits.” The proposed change is necessary to correct a non-conservative TS value for DCPP, Unit 1. The Unit 1 RCS flow specified in TS 3.4.1 for 100 percent power is 359,000 gallons per minute (gpm). However, the TS value is less than the 359,200 gpm RCS minimum measured flow (MMF) value specified in the Updated Final Safety Analyses Report (UFSAR) Table 4.1-1, “Reactor Design Comparison.” The UFSAR RCS MMF value represents the RCS flow value used in the reactor core DNB safety analyses. This issue has been entered in the DCPP corrective action program, and the actual Unit 1 RCS flow value has been verified to be within the limits required by the applicable safety analyses.

In order to resolve the non-conservative TS value, the proposed change would revise the RCS flow requirements in DCPP TS 3.4.1 to be consistent with TS 3.4.1 in NUREG-1431, Revision 4, Volume 1, “Standard Technical Specifications—Westinghouse Plants,” April 2012 (ADAMS Accession No. ML12100A222). The proposed change to the RCS flow requirements in TS 3.4.1 would also be consistent with the NRC-approved Technical Specification Task Force (TSTF) Traveler-339-A, Revision 2, “Relocate TS Parameters to [Core Operating Limits Report] COLR,” and NRC-approved WCAP-14483-A, “Generic Methodology for Expanded Core Operating Limits Report,” dated June 13, 2000 (ADAMS Accession No. ML003723269).

The proposed change would delete the current DCPP, Units 1 and 2 TS 3.4.1 RCS flow Tables 3.4.1-1 and 3.4.1-2, and would add the DCPP, Units 1 and 2 RCS thermal design flow values of 350,800 gpm and 354,000 gpm, respectively, to the requirements of TS 3.4.1. In addition, the proposed change would add the RCS MMF values of 359,200 gpm and 362,500 gpm, to the DCPP, Units 1 and 2 COLR, respectively. Consistent with the Standard Technical Specifications (STS), the proposed change would also include a reference to the RCS COLR flow requirements in the TS 3.4.1 Limiting Condition for Operation and Surveillance Requirements. Due to the elimination of RCS flow Tables 3.4.1-1 and 3.4.1-2, a reference to these tables is also deleted from Figure 2.1.1-1, “Reactor Core Safety Limit.”

As such, the proposed change would resolve the non-conservative TS value for Unit 1 and serve to make the DCPP, Units 1 and 2 TS more consistent with the STS in NUREG-1431.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which 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 proposed change revises the DCPP Unit 1 and Unit 2 RCS flow requirements in TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to be more consistent with TS 3.4.1 in NUREG-1431 and with the applicable DCPP safety analyses. The proposed RCS flow values will ensure the assumptions of the safety analyses continue to be met.

As such, the proposed change does not affect the design or function of any plant structures, systems, and components (SSCs). Thus, the proposed change does not affect plant operation, design features, or any analysis that verifies the capability of an SSC to perform a design function. As the proposed change is consistent with the RCS flow assumptions of the safety analyses, the proposed change does not affect any previously evaluated accidents in the UFSAR. In addition, the proposed change does not affect any SSCs, operating procedures, and administrative controls which have the function of preventing or mitigating any accident previously evaluated in the UFSAR.

The proposed change will not alter any accident analyses assumptions discussed in the UFSAR and will continue to assure the DCPP units operate within the assumptions of the applicable safety analyses described in the UFSAR.

Therefore, the proposed change does 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 accident from any accident previously evaluated?

Response: No.

The proposed change revises the DCPP Unit 1 and Unit 2 RCS flow requirements in TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to be more consistent with TS 3.4.1 in NUREG-1431 and with the applicable DCPP safety analyses. The proposed RCS flow values will ensure the assumptions of the safety analyses continue to be met.

The proposed change does not change any system functions or maintenance activities. The change does not involve physical alteration of the plant, that is, no new or different type of equipment will be installed. The proposed change involves no physical plant modification or changes in plant operation, therefore no new failure modes are created. As such, the proposed change does not create new failure modes or mechanisms that are not identifiable during testing, and no new accident precursors are generated.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. The setpoints at which protective actions are initiated are not altered by the proposed change. Therefore, sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. The proposed RCS flow value changes are consistent with the plant safety analyses. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration.

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, CA 94120.

NRC Branch Chief: Michael T. Markley.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San Diego County, California

Date of amendment request: August 20, 2015. A publicly-available version is in ADAMS under Accession No. ML15236A018.

Description of amendment request: The proposed amendment would revise Appendix 3A of the Updated Final Safety Analysis Report to more fully reflect the permanently shutdown status of the SONGS, Units 2 and 3. The revision would include a limited set of exceptions and clarifications to referenced Regulatory Guides to reflect the significantly reduced decay heat loads in the SONGS, Units 2 and 3 Spent Fuel Pools and to support corresponding design basis changes and modifications that will allow for the implementation of the “cold and dark” strategy outlined in the SONGS Post-Shutdown Decommissioning Activities Report (PSDAR).

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

The only accident previously evaluated, is the Spent Fuel Pool Boiling Event. The initiating event (loss of cooling) would no longer lead to a rapid increase in pool temperature to the boiling point or to a relatively short-term reduction in pool level due to evaporative losses. Currently a loss of cooling would lead to a very slow heat-up toward the boiling point taking at least a week or more. From that point the slower evaporative losses would take several weeks to reduce inventory to unacceptable levels.

The most likely cause of a loss of function of the Spent Fuel Pool Cooling System (SFPCS) is not a failure of components in the cooling system, but instead a loss of electrical power. The probability of a loss of power is substantially higher than the probability of a contemporaneous common cause failure of active components in the cooling system. For example, NRC has collected operating experience on loss of Spent Fuel Pool (SFP) cooling for nuclear plants in the U.S., which includes both safety-related and non-safety-related cooling systems. As indicated in NUREG-1275, Volume 12, the causes of loss of SFP cooling were the loss of the SFP cooling pumps due to loss of electrical power (39 of 56 events), loss of suction from the spent fuel pool, flow blockage, loss of the heat sink, and one case of inadequate configuration control. As concluded by the NRC: “The dominant cause of the actual loss of SFP cooling events was loss of electrical power to the SFP cooling pumps.” There were no cases involving a common cause failure mode, such as seismic events or tornados. Given this operating experience, any increase in the probability of a spent fuel pool boiling event due to the seismic re-classification of the system would be minimal in comparison to the failure rate due to loss of electrical power.

The change in commitment does not affect the consequences of the spent fuel pool boiling accident (which by definition assumes loss of the spent fuel pool cooling system). Revised dose calculations were completed to support the changes to the Updated Final Safety Analysis Report (UFSAR) Chapter 15 Accident Analysis, and the UFSAR was revised to reflect the new analysis. These were recently reviewed to verify they remain bounding for the much slower event, even if it is not terminated (through restored cooling or adequate make-up) prior to reaching levels approaching the top of the stored fuel. This re-evaluation confirmed the doses previously calculated remain bounding and several orders of magnitude below applicable limits.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

The only accident relevant to this proposed change would be an unmitigated Spent Fuel Pool Boiling Event (i.e., boiling without restoration of cooling or make-up prior to uncovering of the spent fuel). The initiating event (loss of cooling) would no longer lead to a rapid increase in pool temperature to the boiling point and a relatively short-term reduction in pool level due to evaporative losses. Currently a loss of cooling would lead to a very slow heatup toward the boiling point taking at least a week or more. From that point the slower evaporative losses would take several weeks to reduce inventory to unacceptable levels. The only safety function remaining relates to maintaining the fuel cladding in the SFP (cooling is not a safety-related function as defined in the updated Chapter 15 Fuel Pool Boiling Accident Analysis, only maintaining water level—Reference 6.12). The only remaining safety related SSCs at SONGS Units 2 and 3 are the Spent Fuel Pool and related structural components (pool liner, structure, and racks).

The Make-up System will ensure that sufficient water is supplied to the SFPs in the event of loss of cooling. In addition to the Seismic Category I make-up source, currently there are numerous other diverse sources of make-up for the SFPs, including:

• As provided in SONGS Units 2 and 3 procedures, the Nuclear Service Water connections located on the SFP operating level can be used via hoses to fill the pool. These connections are QC III, Seismic Category II.
• As provided in SONGS Units 2 and 3 Mitigation Strategies, water from Fire Water Tanks T-102 and T-103 via Fire Pumps P-220 (diesel driven), P-221 or P-222 (both of which are motor driven) can be provided through the installed fire system piping to two fire hose cabinets located on the Spent Fuel Pool Operating level. The tanks, pumps and piping are QC III-EPS and Seismic Category II.
• As provided in SONGS Units 2 and 3 Mitigation Strategies, make-up to the SFPs can be provided using water from one or more of the following sources: Demineralized Water Tanks T-266, T-267 or T-268, all are located at a higher elevation at the Make-up Demineralizer Area at the south end of the plant. Skid mounted pump P-i1058 delivers water from these sources to the seismic standpipe and from the standpipe to the SFP. T-266, T-267 and T-268 are QC III, Seismic Category II. P-1058 is QC III-EPS and Seismic Category III.
• As discussed in SONGS Units 2 and 3 Mitigation Strategies, the 10″ City Water Line Supply Line can be used as an alternate source of SFP make-up water.
• Another make-up path is available using the Seismic Category I Demineralized Water Storage Tank (T-351) located in the North Industrial Area along with Seismic Category I portable diesel driven Fire Pump (P-i1065) using strategically staged hoses between the tank, pump, Seismic Category I standpipe and the Spent Fuel Pool. The hoses are pressure tested annually and are inspected for location quarterly per SONGS Units 2 and 3 procedures.

The Mitigation Strategies are sequenced to assure the strategies can be deployed in 2 hours or less. The capability to achieve this time requirement was evaluated in a formal study and further demonstrated in the field using actual staff, procedures and equipment.

Given the number and diversity of make-up sources, and the time available to supply make-up to the SFPs in the loss of spent fuel pool cooling, it is not credible to postulate a complete loss of make-up to a SFP. As discussed in NRC's June 30, 2014, letter concerning San Onofre Nuclear Generating Station, Units 2 and 3—Rescission of Order EA-12-049:

[T]he time to boil off water inventory in the SFP to a level of 10 feet above the spent fuel will be sufficiently long to obviate the need for additional strategies to restore SFP cooling. The NRC staff concludes that given the low decay heat levels and the long time to boil off, the reliance on the SFP inventory for passive cooling provides an equivalent level of protection as that which would be provided by the initial phase of the guidance and strategies for maintaining or restoring SFP cooling capabilities that would be necessary for compliance with Order EA-12-049 using installed equipment. The staff further concludes that the long time to boil off the SFP inventory to a point at which make-up would be necessary for radiation shielding purposes obviates the need for the transition phase of the guidance and strategies that would be necessary for compliance with Order EA-12-049 using on-site portable equipment. The staff also concludes that the low decay heat and long boil-off period provides sufficient time for the licensee to obtain off-site resources on an ad hoc basis to sustain the SFP cooling function indefinitely, obviating the need for the final phase of the guidance and strategies that would be necessary for compliance with Order EA-12-049.

Similarly, as described in NRC's 2015 exemption from certain emergency planning requirements for SONGS Units 2 and 3:

Additionally, in its letters to the NRC dated October 6, 2014, and December 15, 2014, SCE described the SFP make-up strategies that could be used in the event of a catastrophic loss of SFP inventory. The multiple strategies for providing make-up water to the SFP include: Using existing plant systems for inventory make-up; an internal strategy that relies on installed fire water pumps and service water or fire water storage tanks; or an external strategy that uses portable pumps to initiate make-up flow into the SFPs through a seismic standpipe and standard fire hoses routed to the SFPs or to a spray nozzle. These strategies will continue to be required as a license condition. Considering the very low probability of beyond-design-basis accidents affecting the SFP, these diverse strategies provide defense-in-depth and time to provide additional make-up or spray water to the SFP before the onset of any postulated off-site radiological release.

It is not necessary to postulate both a loss of spent fuel pool cooling in conjunction with a loss of spent fuel pool make-up, and such an event is not postulated in UFSAR Section 15.7.3.8 related to SFP boiling and is not credible given the number of diverse sources of make-up and the time available to supply make-up.

As currently discussed in UFSAR 9.1.2.3, spent fuel pool boiling also will not adversely affect the integrity of the SFPs. The reinforced concrete temperature differences and gradients were determined based on an inside face temperature of 230 °F (water temperature of 212 °F and gamma heating of 18 °F). That analysis indicates that the SFP walls have sufficient structural capability to accommodate this thermal loading.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

The proposed changes do not alter any design basis or safety limits for the plant. The applicable limits are spent fuel clad temperature and spent fuel pool level. The spent fuel cladding temperature is assured by maintaining water level to support natural circulation cooling within the spent fuel racks. Forced cooling keeps evaporative losses and Fuel Handling Building environs within nominal limits. Thus, the SSCs that support the design and safety limits are limited to those that maintain inventory (Spent Fuel Pool and related structural components (pool liner, structure, and racks) and sufficient equipment to replace evaporative or other losses. Complete loss of make-up is not credible given the existence of numerous sources of make-up and the time available to provide make-up. No changes to the pool and its structures are proposed and make-up capability remains assured.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Walker A. Matthews, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, CA 91770.

NRC Branch Chief: Bruce Watson.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, GA

Date of amendment request: August 4, 2015. A publicly-available version is in ADAMS under Accession No. ML15216A602.

Description of amendment request: The licensee describes the application as follows: “This amendment corrects an obvious typographical error in the Unit 1 FOL [Facility Operating License], and on page 5.0.17 of the Unit 2 TS [Technical Specification]. The Degraded Voltage Protection license condition in Part 2.C of the Unit 1 FOL (DPR-57) is currently listed as condition number 10, whereas it should be listed as condition number 11. In addition, this paragraph should be further indented to the right, to clarify that it's a third level paragraph (i.e. level 2.C.11). In addition to the FOL change, this amendment corrects an incorrect Unit number in Hatch Unit 2 TS page 5.0.17. This page was inadvertently sent and issued stating Unit 1 on the bottom left, whereas it should clearly state Unit 2. Lastly, this amendment adds the term STAGGERED TEST BASIS to the Definitions section of the Unit 1 and Unit 2 TS. This term was removed from the TS and moved to the Surveillance Frequency Control Program (SFCP) when the NRC issued the TSTF-425 license amendment in [January 3,] 2012 to relocate specific surveillance frequency requirements to a licensee controlled program. This term, however, was reintroduced into Section 5 of the TS as a defined term when Hatch adopted the Control Room Envelope Habitability Program (TSTF-448) [in an amendment issued on August 29, 2014]. Since it's currently used as a defined term in Section 5 of the TS, it needs to be included in the Definitions section of the TS.”

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment contains no technical changes; all proposed changes are administrative. These changes are consistent with the intent of what has already been approved by the Nuclear Regulatory Commission (NRC).

There are no accidents affected by this change, and therefore no increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment contains no technical changes; all proposed changes are administrative. These changes are consistent with the intent of what has already been approved by the Nuclear Regulatory Commission (NRC).

There are no accidents affected by this change, and therefore no possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment contains no technical changes; all proposed changes are administrative. These changes are consistent with the intent of what has already been approved by the Nuclear Regulatory Commission (NRC).

There are no accidents affected by this change, and therefore no reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, 40 Inverness Center Parkway, Birmingham, AL 35201.

NRC Branch Chief: Robert J. Pascarelli.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated.

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 51.22(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation, and/or Environmental Assessment as indicated. All of these items can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document.

Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power Station, Kewaunee County, Wisconsin

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: November 17, 2014, as supplemented by letter dated August 13, 2015.

Brief description of amendments: The amendments revised the Cyber Security Plan (CSP) Milestone 8 full implementation date as set forth in the CSP Implementation Schedule for the following plants: Kewaunee Power Station; Millstone Power Station, Unit Nos. 2 and 3; North Anna Power Station, Unit Nos. 1 and 2; and Surry Power Station, Unit Nos. 1 and 2.

Date of issuance: October 7, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 216, 323, 269, 276, 258, 286, and 286. A publicly-available version is in ADAMS under Accession No. ML15245A482. Documents related to these amendment are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-43, DPR-65, DPR-49, NPF-4, NPF-7, DPR-32, and DPR-37: Amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: May 5, 2015 (80 FR 25718). The supplement letter dated August 13, 2015, 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.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 7, 2015.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

Date of amendment request: September 10, 2015, as supplemented by letters dated September 30 and October 20, 2015.

Brief description of amendment: The amendment approved a one-time extension of the Technical Specification (TS) completion time associated with the Division 2 Shutdown Service Water Subsystem from 72 hours to 7 days in support of maintenance activities.

Date of issuance: October 22, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment No: 207. A publicly-available version is in ADAMS under Accession No. ML15280A258; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-62: The amendment revised the TSs and License.

Date of initial notice in Federal Register : September 18, 2015 (80 FR 56498).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 22, 2015.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373, LaSalle County Station, Unit 1 and Unit 2, LaSalle County, Illinois

Date of amendment request: January 12, 2015.

Brief description of amendments: The amendments deleted the limiting condition for operation (LCO) Note for Technical Specification (TS) Section 3.5.1, “ECCS [emergency core cooling system]—Operating.” The current Note allowed the licensee to consider the low pressure coolant injection subsystem associated with the residual heat removal system to be OPERABLE under specified conditions.

Date of issuance: October 14, 2015.

Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.

Amendment Nos.: 217 and 203. A publicly-available version is in ADAMS under Accession No. ML15244B410; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-11 and NPF-18: Amendments revised the Facility Operating License and TSs.

Date of initial notice in Federal Register: March 31, 2015 (80 FR 17091).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 14, 2015.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: December 19, 2014, as supplemented by letter dated June 26, 2015.

Brief description of amendment: This amendment revised the technical specifications (TSs) to adopt performance-based Type C testing for the reactor containment, which would allow for extended test intervals for Type C valves, and corrects an editorial issue in the TSs.

Date of issuance: October 9, 2015.

Effective date: As of the date of issuance and shall be implemented within 45 days from the date of issuance.

Amendment No.: 288. A publicly-available version is in ADAMS under Accession No. ML15239B293; documents related to this amendment are listed in the Safely Evaluation enclosed with the amendment.

Facility Operating License No. NPF-3: Amendment revised the Facility Operating License and TSs.

Date of initial notice in Federal Register: March 31, 2015 (80 FR 17090), and July 7, 2015 (80 FR 38759). The supplemental letter dated June 26, 2015, provided additional information that clarified the application, did not expand the scope of the application as previously noticed, and did not change the staff's 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 9, 2015.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: December 30, 2014.

Brief description of amendment: This amendment revises the technical specification (TS) surveillance requirement for the frequency to verify that each containment spray system nozzle is unobstructed from every 10 years to an event-based frequency.

Date of issuance: October 20, 2015.

Effective date: As of the date of issuance and shall be implemented within 45 days from the date of issuance.

Amendment No.: 289. A publicly-available version is in ADAMS under Accession No. ML15251A046; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-3: Amendment revised the Facility Operating License and TSs.

Date of initial notice in Federal Register: March 31, 2015 (80 FR 17090).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 20, 2015.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power Plant, Unit 3, Humboldt County, California

Date of amendment request: June 30, 2014, as supplemented March 27, 2015.

Brief description of amendment: The amendment revised the Humboldt Bay Power Plant, Unit 3 License to approve the revised Emergency Plan.

Date of issuance: September 23, 2015.

Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.

Amendment No.: 46. A publicly-available version is in ADAMS under Accession No. ML15148A361; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. DPR-7: Amendment revised the Facility Operating License.

Date of initial notice in Federal Register: August 19, 2014 (79 FR 49109).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 23, 2015.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, Fairfield County, South Carolina

Date of amendment request: July 22, 2015.

Brief description of amendment: The amendment revised Technical Specification Section 6.0, “Administrative Controls,” by changing the “Shift Supervisor” title to “Shift Manager.”

Date of issuance: October 15, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment No.: 202. A publicly-available version is in ADAMS under Accession No. ML15208A029; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-12: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: August 14, 2015 (80 FR 48924), as corrected by Federal Register notice dated August 20, 2015 (80 FR 50663).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 15, 2015.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee

Date of amendment request: June 17, 2015, as supplemented by letters dated July 14, August 3, August 28, September 3, and September 21, 2015.

Brief description of amendment: The amendment adopted new Technical Specification (TS) 3.7.16, “Component Cooling System (CCS)—Shutdown,” and TS 3.7.17, “Essential Raw Cooling Water (ERCW) System—Shutdown,” and revised TS 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” and TS 3.4.6, “RCS Loops-MODE 4,” to support dual-unit operation of WBN Units 1 and 2.

Date of issuance: October 20, 2015.

Effective date: As of the date of issuance and shall be implemented after the issuance of the Facility Operating License for Unit 2.

Amendment No.: 104. A publicly-available version is in ADAMS under Accession No. ML15275A042; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment.

Facility Operating License No. NPF-90: Amendment revised the Facility Operating License and TSs.

Date of initial notice in Federal Register: July 17, 2015 (80 FR 42552). The supplemental letters dated July 14, August 3, August 28, September 3, and September 21, 2015, provided additional information that clarified the application. These supplements did not change the staff's proposed no significant hazards consideration. The supplemental letter dated September 3, 2015, provided additional information that expanded the scope of the application as originally noticed. A notice published in the Federal Register on September 15, 2015 (80 FR 55383), supersedes the original notice in its entirety to update the expanded scope of the amendment description and include the staff's proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in an SE dated October 20, 2015.

No significant hazards consideration determination comments received: No.

Dated at Rockville, Maryland, this 2nd day of November, 2015.

For the Nuclear Regulatory Commission.

Anne T. Boland,

Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.