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

Background

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission 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 July 12, 2012 to July 25, 2012. The last biweekly notice was published on July 24, 2012 (77 FR 43374).

ADDRESSES:

You may access information and comment submissions related to this document, which the NRC possesses and are publicly available, by searching on http://www.regulations.gov under Docket ID NRC-2012-0181. You may submit comments by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0181. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-3668; email: Carol.Gallagher@nrc.gov.
• Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
• Fax comments to: RADB at 301-492-3446.

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

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2012-0181 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and are publicly available, by any of the following methods:

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0181.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library 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 pdr.resource@nrc.gov. Documents may be viewed in ADAMS by performing a search on the document date and docket number.
• 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-2012-0181 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

The NRC cautions you not to include identifying or contact information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS, and the NRC does not 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 in their comment submissions that they do not want to be publicly disclosed. Your request should state that the NRC will not edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.92, 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; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. 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.

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 “Rules of Practice for Domestic Licensing Proceedings” 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 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 by the above date, 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 identify 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.

If a hearing is requested, 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.

All documents filed in the 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 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 hearing.docket@nrc.gov, 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/apply-certificates.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 the 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 agency'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 Web site at http://www.nrc.gov/site-help/e-submittals.html,, by email at MSHD.Resource@nrc.gov, 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. 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. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov.

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South Carolina

Date of amendment request: June 8, 2012.

Description of amendment request: The proposed change would revise the Technical Specifications (TSs) 3.1.4, “Rod Group Alignment Limits,” and TS 3.1.7, “Rod Position Indication,” to allow up to 1 hour of soak time following substantial rod movement during which individual rod position indicators may not be within its limits.

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.

This license amendment request proposes to allow up to one hour of soak time following substantial rod movement during which time the rod position indication may be outside its limits. This would allow an additional hour for rod position indication to be inoperable or a control rod to be misaligned prior to entry into a TS LCO [Limiting Condition for Operation] Condition and Required Actions. RPI [Rod Position Indicators] instrumentation is not an assumed accident initiator; however, the HBRSEP, Unit No. 2 safety analyses consider two types of rod misalignment events, static misalignment and a dropped rod.

The safety analyses show that for the static misalignment event, without any operator intervention, a single fully withdrawn rod event does not result in any fuel pin failure; therefore, the static rod misalignment event is not time dependent and an additional hour, with the misalignment undetected and unmitigated does not increase the consequences of the event. Multiple rod misalignment events are bounded by the single rod misalignment analyses and therefore an additional hour would not have any impact on this event.

The safety analyses also show that a single dropped rod event, without any operator intervention, does not result in any fuel pin failure; therefore, the rod drop event is not time dependent and an additional hour with the misalignment undetected and unmitigated does not increase the consequences of the event. Multiple rod drop events cause the reactor to trip and therefore an additional hour would not have any impact on that event.

Although this license amendment request may allow a misaligned rod to be undetected for an additional hour, the additional time for discovery does not change the probability of a misaligned control rod event because the one hour time extension does not affect the control rod drive system features that would result in either type of misalignment.

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 previously evaluated?

Response: No.

This proposed change does not alter the design, function, or operation of any plant component and does not install any new or different equipment. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of these changes. No new equipment performance burdens are imposed.

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 the margin of safety?

Response: No.

The RPI system is an instrumentation system that provides indication to the operators that a control rod may be misaligned. Inoperable individual RPI instrumentation does not, by itself in any way, harm or impact reactor operation. Inoperable rod position indication may impair the ability of the operators to detect a misaligned rod. However, the impact of inoperable RPI instrumentation may be offset by availability of other indications that a rod is misaligned such as nuclear instrumentation indication that reactor power has shifted to one side of the core or thermocouple indication that the core temperatures increased in one region of the core and/or decreased in another region of the core. Based on plant experience, the likelihood of a misaligned rod at HBRSEP, Unit No. 2 is considered to be small and the likelihood of a misaligned rod coincident with inoperable rod position indication during the allowed one hour extension is even smaller. In addition, these proposed changes may enhance plant safety and reliability because the one hour soak time will allow the operators and engineers to focus on monitoring the reactor performance without unnecessary entry into TS LCO Conditions and Required Actions.

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: David T. Conley, Manager—Senior Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Acting Branch Chief: Jessie F. Quichocho.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

Date of amendment request: April 30, 2012.

Description of amendment request: This amendment request proposes to permanently revise technical specification (TS) 6.8.4.j, Steam Generator (SG) Surveillance Program, to exclude portions of the SG tube below the top of the SG tubesheet from periodic tube inspections.

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 previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed change that alters the SG inspection and reporting criteria does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident.

Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed change to the SG tube inspection and repair criteria are the SG tube rupture (SGTR) event and the steam line break (SLB) postulated accident.

Addressing the SGTR event, the required structural integrity margins of the SG tubes and the tube-to-tubesheet joint over the H* distance will be maintained. Tube rupture in tubes with cracks within the tubesheet is precluded by the constraint provided by the presence of the tubesheet and the tube-to-tubesheet joint. Tube burst cannot occur within the thickness of the tubesheet. The tube-to-tubesheet joint constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet, and from the differential pressure between the primary and secondary side, and tubesheet rotation. The structural margins against burst, as discussed in Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR [Pressurized-Water Reactors] Steam Generator Tubes” [Reference 7] and NEI [Nuclear Energy Institute] 97-06, “Steam Generator Program Guidelines”, [Reference 3] are maintained for both normal and postulated accident conditions.

For the portion of the tube outside of the tubesheet, the proposed change also has no impact on the structural or leakage integrity. Therefore, the proposed change does not result in a significant increase in the probability of the occurrence of a SGTR accident.

At normal operating pressures, leakage from primary water stress corrosion cracking below the proposed limited inspection depth is limited by the tube-to-tubesheet crevice. Consequently, negligible normal operating leakage is expected from degradation below the inspected depth within the tubesheet region. The consequences of an SGTR event are not affected by the primary to secondary leakage flow during the event as primary to secondary leakage flow through a postulated tube that has been pulled out of the tubesheet is essentially equivalent to a tube rupture. Therefore, the proposed change does not result in a significant increase in the consequences of an SGTR. In addition, the selected H* value envelopes the depth within the tubesheet required to prevent a tube pullout.

The probability of a SLB is unaffected by the potential failure of a SG tube as the failure of a tube is not an initiator for a SLB event.

The leak rate factor of 1.82 for Turkey Point Units 3 and 4, for a postulated SLB, has been calculated as shown in References 2, 9 and 19. Turkey Point Units 3 and 4 will apply the factor of 1.82 to the normal operating leakage associated with the tubesheet expansion region in the condition monitoring (CM) and operational assessment (OA). Through application of the limited tubesheet inspection scope, the existing operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) will not occur. Multiplying the TS operational leak rate limit of 150 gpd (at room temperature) through any one SG by a factor of 1.82 shows that the maximum primary to secondary accident induced leak rate is limited to 273 gpd. This leakage rate is bounded by the current licensing basis assumed primary to secondary accident leak rate of 0.20 gpm (288 gpd) through any one SG for SLB. Since the existing limit on operational leakage continues to ensure that the SLB assumed accident induced leakage will not be exceeded, the consequences of a SLB accident are not increased.

For the CM assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 1.82 and added to the total leakage from any other source and compared to the allowable accident induced leak rate. For the OA, the difference in the leakage between the allowable leakage and the calculated accident induced leakage from sources other than the tubesheet expansion region will be divided by 1.82 and compared to the observed operational leakage.

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

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

Response: No

The proposed change that alters the SG inspection and reporting criteria does not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures. Plant operation will not be altered, and all safety functions will continue to perform as previously assumed in accident analyses. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No

The proposed change defines the safety significant portion of the tube that must be inspected and repaired. WCAP-17345, Rev. 2 [Reference 9] identifies the specific inspection depth below which any type of tube degradation is shown to have no impact on the performance criteria in NEI 97-06 Rev. 3, “Steam Generator Program Guidelines” [Reference 3] and TS 6.8.4.j, “Steam Generator (SG) Program.”

The proposed change that alters the SG inspection and reporting criteria maintains the required structural margins of the SG tubes for both normal and accident conditions. Nuclear Energy Institute 97-06, “Steam Generator Program Guidelines” [Reference 3], and NRC Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR Steam Generator Tubes” [Reference 7], are used as the bases in the development of the limited tubesheet inspection depth methodology for determining that SG tube integrity considerations are maintained within acceptable limits. RG 1.121 describes a method acceptable to the NRC for meeting General Design Criteria (GDC) 14, “Reactor Coolant Pressure Boundary,” GDC 15, “Reactor Coolant System Design,” GDC 31, “Fracture Prevention of Reactor Coolant Pressure Boundary,” and GDC 32, “Inspection of Reactor Coolant Pressure Boundary,” by reducing the probability and consequences of a SGTR. RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation, the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking, Westinghouse WCAP-17091-P, Rev. 0 [Reference 2] and WCAP-17345, Rev. 2 [Reference 9] define a length of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary to secondary leakage during all plant conditions. The SLB leak rate factor for Turkey Point Units 3 and 4 is 1.82 (Table 9-7 in WCAP-17091-P). Multiplying the TS operational leak rate limit of 150 gpd through any one SG by the leak rate factor of 1.82 shows that the maximum primary to secondary accident induced leak rate is limited to 273 gpd. This leakage rate is bounded by the current licensing basis assumed primary to secondary accident leak rate of 0.20 gpm (288 gpd) through any one SG for SLB.

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

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: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Acting Branch Chief: Jessie F. Quichocho.

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

Date of amendment request: May 30, 2012.

Description of amendment request: The proposed amendment would revise Technical Specification Section 2.0, “Safety Limits.” Specifically, the proposed amendment would revise two recirculation loop and single recirculation loop Safety Limit Minimum Critical Power Ratio (SLMCPR) values to reflect results of a cycle-specific calculation.

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?

Response: No.

Four accidents have been evaluated previously as reflected in the CNS [Cooper Nuclear Station] Updated Safety Analysis Report (USAR). These four accidents are (1) loss-of-coolant, (2) control rod drop, (3) main steam line break, and (4) fuel handling. The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. Changing the SLMCPR values does not increase the probability of an evaluated accident. The change does not require any physical modifications to the plant or any components, nor does it require a change in plant operation. Therefore, no individual precursors of an accident are affected.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. This proposed change makes no modification to the design or operation of the systems that are used in mitigation of accidents. Limits have been established, consistent with Nuclear Regulatory Commission (NRC) approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change to the values of the SLMCPR continues to conservatively establish this safety limit such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.

Based on the above, NPPD [Nebraska Public Power District] concludes that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

Creation of the possibility of a new or different kind of accident from an accident previously evaluated would require creation of precursors of that accident. New accident precursors may be created by modification of the plant configuration or changes in how the plant is operated. The proposed change does not involve a modification of the plant configuration or in how the plant is operated. The proposed change to the SLMCPR values assures that safety criteria are maintained.

Based on the above, NPPD concludes that the proposed change does 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?

Response: No.

The values of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1% of fuel rods are expected to be in boiling transition if the Minimum Critical Power Ratio limit is not violated. The proposed change will ensure the appropriate level of fuel protection is maintained. Additionally, operational limits are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria are met (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences).

Based on the above, NPPD concludes that 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. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit 1, Washington County, Nebraska

Date of amendment request: February 10, 2012.

Description of amendment request: The proposed amendment would establish the limiting condition for operation (LCO) requirements for the reactor protective system (RPS) actuation circuits in Technical Specification (TS) 2.15, “Instrumentation and Control Systems.” Specifically, the proposed change: renumbers LCO 2.15(1) through 2.15(4) to 2.15.1(1) through 2.15.1(4), renumbers LCO 2.15(5) to LCO 2.15.3 with an associated Table 2-6, and implements a new LCO 2.15.2 for the RPS logic and trip initiation channels. The Table of Contents will also be revised to reflect the renumbering and addition of the LCO for the RPS logic and trip initiation channels and the new Table 2-6.

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 reactor protective system logic and trip initiation channels meets Criterion 3 of 10 CFR 50.36 for inclusion into Technical Specification (TS) as a component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient. The TSs currently does not have limiting conditions for operations (LCO) specific for this circuitry, but does contain surveillance requirements. The addition of LCOs provides additional restrictions on the operation of the plant and provides required actions and time limits if these components are incapable of performing their function. As such, the proposed change does not increase the probability of an accident. The proposed changes do not alter the physical design of the RPS, or any other plant structure, system or component (SSC) at Fort Calhoun Station (FCS).

The proposed changes conform to the Nuclear Regulatory Commission's (NRC's) regulatory guidance regarding the content of plant TS as identified in 10 CFR 50.36 and NRC publication NUREG 1432.

Therefore, 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 TS changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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

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

Response: No.

The TS operability requirements for the RPS logic and trip initiation channels ensure there is adequate components operable to assure safe reactor operation and are necessary to ensure safety systems accomplish their safety function for design basis accident events. The proposed TS would revise the applicability for when the RPS logic and trip initiation channels are required to be operable to include whenever control element assemblies (CEAs) are capable of being withdrawn and the reactor coolant system (RCS) is not at refueling boron concentration. When the RCS boron concentration is at refueling boron concentration, or when no more than one trippable control rod is capable of being withdrawn, the RPS function is already fulfilled. These proposed TS changes for the RPS are aligned with the applicability and operability requirements provided in NUREG 1432.

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: David A. Repka, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Michael T. Markley.

ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station (Zion), Units 1 and 2, Lake County, Illinois

Date of amendment request: May 31, 2012.

Description of amendment request: The proposed amendments would approve methods of analysis, use of the upgraded fuel handling building crane system as a single-failure proof crane, and a NUREG 0612 compliant heavy loads handling program.

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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The existing DSAR [Defueled Safety Analysis Report] analysis assumes that a spent fuel cask drop occurs. In this analysis, the physics of the drop, coupled with concrete bumpers on the cask loading pit and pool edge were used to demonstrate that a postulated drop of the spent fuel cask near the Spent Fuel Pool neither impacted the spent fuel directly nor damaged the pool structure in a manner that adversely affected the spent fuel, when a cask was to be handled in the cask loading pit. The proposed License Amendment Request to operate a single-failure proof Fuel Building Crane demonstrates that no analysis is required for the cask drop event based on the design and the associated programmatic controls. A drop of the spent fuel cask handled with a single-failure proof crane (designed to ASME NOG-1 [“Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder)”] and compliant with NUREG-0554 [“Single-Failure-Proof Cranes for Nuclear Power Plants”, ML110450636]), operated in accordance with the administrative controls of NUREG-0612 [“Control of Heavy Loads at Nuclear Power Plants,” ML070250180] has an acceptably low probability so as to effectively preclude consideration of the event. The risk of such a drop event using the new single-failure proof crane operated in accordance with the Heavy Loads Program procedures, qualitatively, is lower than the event previously analyzed which postulate the event without evaluation of its likelihood.

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

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

Response: No.

The location and design functions of the Fuel Building crane are not changed from those currently described in the DSAR. Because the new crane has a single-failure proof design the uncontrolled lowering, or drop, of a heavy load will not be considered credible. Evaluations show that individual malfunctions or component failures of the crane will not result in load drop. The new single-failure proof crane['s] primary use[s] will be to move a loaded or unloaded MAGNASTOR transfer cask between the cask loading pit [and] the decontamination pit, and transfer [the cask] to the low profile cart rail transport in the Fuel Handling Building. No components that are classified as Important to the Defueled Condition, other than the Fuel Building crane, will be affected by these movements. Based on the design and programmatic controls on the crane, no load will lower uncontrollably or drop in or around the spent fuel pool or near an open cask containing spent fuel nor will a cask containing spent fuel drop or be lowered uncontrollably during operation of the crane. Hence no new accidents will be initiated.

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

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

Response: No.

This proposed License Amendment Request involves the replacement of the existing non-single-failure proof Fuel Building Crane with a new single-failure proof crane. The new crane has been designed to meet the specifications found in ASME NOG-1-2004, which has been endorsed by the NRC in RIS 2005-25, as supplemented, as an acceptable means of meeting the criteria in NUREG-0554, “Single-failure Proof Cranes for Nuclear Power Plants.” to provide adequate protection and safety margin against the uncontrolled lowering of the lifted load. The occurrence of a cask load drop accident is considered not credible when the load is lifted with a single-failure proof lifting system meeting the guidance in NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants” Section 5.1.6, “Single-Failure-Proof Handling Systems.” As a result, the proposed change, replacing the existing non-single-failure proof crane, has no adverse impact on stored spent fuel, or structural integrity of the pool.

The configuration of the crane and the primary load, a spent fuel cask containing spent fuel, is changed from that of the DSAR. The specific analysis dealing with a drop of the cask will no longer be applicable and [will be] removed from the DSAR, since the new single-proof crane makes that event of low enough probability to not be considered credible. The maximum critical lift capacity of the crane has not been changed, though the load to be lifted is larger. The structural analyses of the crane and its support structure, however, show acceptable margin under the acceptance criteria of NOG-I for operation of the crane.

Therefore, the proposed change does not involve a significant reduction in the 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: Russ Workman, Deputy General Counsel, EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 84101.

NRC Branch Chief: Bruce Watson.

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 are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through the Agencywide Documents Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html . If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr.resource@nrc.gov.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: July 21, 2011.

Brief description of amendments: The amendments revised Technical Specifications 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” 3.5.4, “Refueling Water Storage Tank (RWST),” and 3.6.6, “Containment Spray System.”

Date of issuance: July 25, 2012.

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

Amendment Nos.: Unit 1-269 and Unit 2-265.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: March 20, 2012 (77 FR 16274).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 25, 2012.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Units 2 and 3 (IP2 and IP3), Westchester County, New York

Date of application for amendment: July 8, 2009, as supplemented by letters dated September 28, 2009, October 26, 2009, October 5, 2010, October 28, 2010, July 28, 2011, August 23, 2011, October 28, 2011, December 15, 2011, January 11, 2012, March 2, 2012, April 23, 2012, and May 7, 2012.

Brief description of amendment: The amendment authorizes the transfer of spent fuel from the IP3 spent fuel pool to the IP2 spent fuel pool, using a newly-designed shielded transfer canister, for further transfer to the on-site Independent Spent Fuel Storage Installation.

Date of issuance: July 13, 2012.

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

Amendment No.: 268 and 246.

Facility Operating License Nos. DPR-26 and DPR-64: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: January 21, 2010 (75 FR 3497).

The supplements provided additional information that clarified the application but did not expand the scope of the application as originally noticed.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 13, 2012.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of application for amendment: September 8, 2010, as supplemented by letters dated November 18, 2010, November 23, 2010, February 23, 2011 (four letters), March 9, 2011 (two letters), March 22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April 21, 2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June 15, 2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011, August 25, 2011, August 29, 2011, August 30, 2011, September 2, 2011, September 9, 2011, September 12, 2011, September 15, 2011, September 26, 2011, October 10, 2011, October 24, 2011, November 14, 2011, November 25, 2011, November 28, 2011, December 19, 2011, February 6, 2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21, 2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May 9, 2012, and June 12, 2012.

Brief description of amendment: The amendment increased the maximum steady-state reactor core power level from 3,898 megawatts thermal (MWt) to 4,408 MWt, which is an increase of approximately 15 percent from the original licensed thermal power level of 3,833 MWt. The proposed increase in power level is considered an extended power uprate.

Date of issuance: July 18, 2012.

Effective date: As of the date of issuance and shall be implemented within 120 days of issuance.

Amendment No: 191.

Facility Operating License No. NPF-29: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: January 10, 2011 (76 FR 1464). The supplemental letters dated November 18, 2010, November 23, 2010, February 23, 2011 (four letters), March 9, 2011 (two letters), March 22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April 21, 2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June 15, 2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011, August 25, 2011, August 29, 2011, August 30, 2011, September 2, 2011, September 9, 2011, September 12, 2011, September 15, 2011, September 26, 2011, October 10, 2011, October 24, 2011, November 14, 2011, November 25, 2011, November 28, 2011, December 19, 2011, February 6, 2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21, 2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May 9, 2012, and June 12, 2012, 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 July 18, 2012.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410, Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

Date of application for amendments: July 20, 2011, as supplemented on November 3, 2011, and January 12, 2012.

Brief description of amendments: The amendments revised the NMP1 Technical Specification (TS) Section 5.1, “Site,” and associated TS Figure 5.1-1, “Site Boundaries, Nine Mile Point-Unit 1,” and the NMP2 TS Figure 4.1-1, “Site Area and Land Portion of Exclusion Area Boundaries,” to reflect the transfer of a portion of the Nine Mile Point Nuclear Station, LLC (NMPNS) site real property located outside of the NMPNS Protected Area but within the current NMPNS Owner Controlled Area, as well as specified easements over the remainder of the NMPNS site, to Nine Mile Point 3 Nuclear Project, LLC (NMP3), a subsidiary of UniStar Nuclear Energy, LLC.

Date of issuance: July 12, 2012.

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

Amendment Nos.: 212 for Unit 1 and 142 for Unit 2.

Renewed Facility Operating License Nos. DPR-63 and NPF-69: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: December 27, 2011 (76 FR 80977).

The supplements dated November 3, 2011, and January 12, 2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's initial proposed no significant hazards consideration determination noticed in the Federal Register on December 27, 2011 (76 FR 80977).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 12, 2012.

No significant hazards consideration comments received: No.