[Federal Register Volume 76, Number 95 (Tuesday, May 17, 2011)]
[Notices]
[Pages 28470-28479]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-11804]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0104]


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

I. 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 
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 April 21, 2011 to May 4, 2011. The last 
biweekly notice was published on May 3, 2011 (76 FR 24926).

Notice of Consideration of Issuance of Amendments to Facility Operating 
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

[[Page 28471]]

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.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852.
    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. 
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 Commission's PDR, located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, 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 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 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
[email protected], 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

[[Page 28472]]

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 
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 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 e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the document to the NRC 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 Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
[email protected], 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 Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible from the 
ADAMS Library online 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 PDR 
Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to 
[email protected].

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of amendment request: March 7, 2011.
    Description of amendment request: The proposed amendment would add 
an applicability period of 42.1 effective full power years (EFPY) to TS 
LCO 3.4.3, figures 3.4.3-1 and 3.4.3-2 which contain the pressure-
temperature (P/T) limit curves for primary coolant system (PCS) heatup 
and cooldown, and limiting condition for operation (LCO) 3.4.12 figure 
3.4.12-1, which contains the low temperature overpressure protection 
(LTOP) setpoint limit curve.
    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.
    No changes are being made to the existing pressure-temperature 
(P/T) limit curves in TS Limiting Condition for Operation (LCO) 
3.4.3

[[Page 28473]]

Figures 3.4.3-1 and 3.4.3-2 and the low temperature overpressure 
(LTOP) setpoint limit curve in LCO 3.4.12 Figure 3.4.12-1. The P/T 
limits curves and the LTOP setpoint limit curve are only being 
revised to add the applicability period of 42.1 effective full power 
years. This applicability period has been verified to be 
conservative for operation through the expiration of the operating 
license on March 24, 2031.
    The changes to the TS figures are applicable to normal plant 
operations and do not influence the probability of occurrence or 
safety analysis considerations for design basis accidents. 
Consequently, there will be no change to the probability or 
consequences of accidents previously evaluated. Operating the 
facility in accordance with the P/T limit and LTOP setpoint limit 
curves ensures that stresses caused by the thermal gradient through 
the RV beltline material remain bounded by the stress analyses. The 
proposed amendment does not involve operation of required 
structures, systems, or components in a manner or configuration 
different than previously recognized or evaluated. No radiological 
barriers are affected by the change.
    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.
    No changes are being made to the existing P/T limit curves in TS 
Figures 3.4.3-1 and 3.4.3-2 and or in the existing LTOP setpoint 
limit curves in TS Figure 3.4.12-1. The TS figures are only being 
changed to add the applicability period of 42.1 effective full power 
years for the P/T limits and LTOP setpoint limit curves. Adding the 
applicability periods to the TS figures will not create the 
possibility of any new or different kind of accidents.
    The change does not involve a modification of plant structures, 
systems, or components. The change will not affect the manner in 
which the plant is operated and will not degrade the reliability of 
structures, systems, or components. Equipment protection features 
will not be deleted or modified, equipment redundancy or 
independence will not be reduced, and supporting system performance 
will not be affected. No new failure modes or mechanisms will be 
introduced as a result of this proposed change.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    Appendix G to 10 CFR Part 50 describes the conditions that 
require P/T limits and provides the general bases for these limits. 
Operating limits based on the criteria of Appendix G, as defined by 
applicable regulations, codes and standards, provide reasonable 
assurance that non-ductile or rapidly propagating failure will not 
occur. The P/T limits are prescribed for all plant modes to avoid 
encountering pressure, temperature, and temperature rate of change 
conditions that might cause undetected flaws to propagate and cause 
non-ductile failure of the reactor coolant pressure boundary. 
Calculation of P/T limits in accordance with the criteria of 
Appendix G to 10 CFR Part 50 and applicable regulatory requirements 
ensures that adequate margins of safety are maintained and there is 
no significant reduction in a margin of safety.
    No change is being made to the existing P/T limit curves or LTOP 
setpoint curve. Only the applicability period associated with the P/
T Limits and LTOP setpoints is being extended. Since the P/T limits 
and LTOP setpoint limits remain unchanged there is no reduction in a 
margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no change or impact on any safety 
analysis assumption or on any other parameter affecting the course 
of an accident analysis supporting the basis of any Technical 
Specification. The proposed change does not involve any increase in 
calculated off-site dose consequences.
    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 Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 4, 2011.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to define a new time limit 
for restoring inoperable reactor coolant system (RCS) leakage detection 
instrumentation to operable status; establish alternate methods of 
monitoring RCS leakage when one or more required monitors are 
inoperable and make conforming TS Bases changes. These changes are 
consistent with NRC-approved Revision 3 to Technical Specification Task 
Force (TSTF) Standard Technical Specification (STS) Change Traveler 
TSTF-514, ``Revise BWR Operability Requirements and Actions for RCS 
Leakage Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (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 clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
Reactor Coolant System (RCS) leakage detection instrumentation 
monitor is the drywell atmospheric gaseous radiation monitor. The 
monitoring of RCS leakage is not a precursor to any accident 
previously evaluated. The monitoring of RCS leakage is not used to 
mitigate the consequences of any accident previously evaluated.
    Therefore, it is concluded that this 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 clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radiation monitor. 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.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radiation monitor. Reducing the amount of time 
the plant is allowed to operate with only the drywell atmospheric 
gaseous radiation monitor operable increases the margin of safety by 
increasing the likelihood that an increase in RCS leakage will be 
detected before it potentially results in gross failure.
    Therefore, it is concluded that 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

[[Page 28474]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the requested amendments involve no 
significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Robert D. Carlson.

Indiana Michigan Power Company (the Licensee), Docket Nos. 50-315 and 
50-316, Donald C. Cook Nuclear Plant, Units 1 and 2 (DCCNP-1), Berrien 
County, Michigan

    Date of amendment request: March 18, 2011.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS), removing the specific 
isolation time for the main steam and main feedwater isolation valves 
(MSIVs) from Surveillance Requirements 3.7.2.1, 3.7.3.1, and 3.7.3.2. 
These changes were previously approved generically by the NRC staff and 
are tracked as Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler TSTF-491.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee 
incorporated by reference the no significant hazards consideration 
(NSHC) analysis endorsed by the NRC staff in a December 29, 2006, 
Federal Register notice (71 FR 78472) and which was published in an 
October 5, 2006, Federal Register notice (71 FR 58884). The October 5, 
2006, NSHC analysis is reproduced below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows relocating main steam and main 
feedwater valve isolation times to the Licensee Controlled Document 
that is referenced in the Bases. The proposed change is described in 
Technical Specification Task Force (TSTF) Standard TS Change 
Traveler TSTF-491 related to relocating the main steam and main 
feedwater valves isolation times to the Licensee Controlled Document 
that is referenced in the Bases and replacing the isolation time 
with the phase ``within limits.''
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes relocate the main steam and main feedwater 
isolation valve times to the Licensee Controlled Document that is 
referenced in the Bases. The requirements to perform the testing of 
these isolation valves are retained in the TS. Future changes to the 
Bases or licensee-controlled document will be evaluated pursuant to 
the requirements of 10 CFR 50.59, ``Changes, test and experiments,'' 
to ensure that such changes do not result in more than minimal 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological consequences of any accident previously 
evaluated. Further, the proposed changes do not increase the types 
and the amounts of radioactive effluent that may be released, nor 
significantly increase individual or cumulative occupation/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed changes relocate the main steam and main feedwater 
valve isolation times to the Licensee Controlled Document that is 
referenced in the Bases. In addition, the valve isolation times are 
replaced in the TS with the phase ``within limits.'' The changes do 
not involve a physical altering of the plant (i.e., no new or 
different type of equipment will be installed) or a change in 
methods governing normal plant operation. The requirements in the TS 
continue to require testing of the main steam and main feedwater 
isolation valves to ensure the proper functioning of these isolation 
valves.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed changes relocate the main steam and main feedwater 
valve isolation times to the Licensee Controlled Document that is 
referenced in the Bases. In addition, the valve isolation times are 
replaced in the TS with the phase ``within limits.'' Instituting the 
proposed changes will continue to ensure the testing of main steam 
and main feedwater isolation valves. Changes to the Bases or license 
controlled document are performed in accordance with 10 CFR 50.59. 
This approach provides an effective level of regulatory control and 
ensures that main steam and feedwater isolation valve testing is 
conducted such that there is no significant reduction in the margin 
of safety.
    The margin of safety provided by the isolation valves is 
unaffected by the proposed changes since there continue to be TS 
requirements to ensure the testing of main steam and main feedwater 
isolation valves. The proposed changes maintain sufficient controls 
to preserve the current margins of safety.
    The NRC staff has reviewed the above 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: James M. Petro, Jr., Senior Nuclear Counsel, 
Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: February 23, 2011.
    Description of amendment request: The proposed amendment would 
adopt an approved change to the standard technical specifications (TSs) 
for General Electric Plants, BWR/4 (NUREG-1433), to allow relocation of 
specific TS surveillance frequencies to a licensee controlled program. 
The proposed change is described in Technical Specification Task Force 
(TSTF) Traveler, TSTF-425, Revision 3 (Rev. 3) (ADAMS Accession No. 
ML090850642) related to the Relocation of Surveillance Frequencies to 
Licensee Control-RITSTF (Risk-Informed TSTF) Initiative 5b and was 
described in the Notice of Availability published in the Federal 
Register on July 6, 2009 (74 FR 31996).
    The proposed change is consistent with NRC-approved Industry/
Technical Specification Task Force Traveler, TSTF-425, Rev. 3, 
``Relocate Surveillance Frequencies to Licensee Control-RITSTF 
Initiative 5b.'' The proposed change relocates surveillance frequencies 
to a licensee-controlled program, the Surveillance Frequency Control 
Program (SFCP). This change is applicable to licensees using 
probabilistic risk guidelines contained in NRC-approved NEI 04-10, 
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed 
Method for Control of Surveillance Frequencies,'' (ADAMS Accession No. 
ML071360456).
    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 any accident previously 
evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic

[[Page 28475]]

surveillance requirements to licensee control under a new program--
the SFCP. Surveillance frequencies are not an initiator to any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased. The 
systems and components required by the Technical Specifications (TS) 
for which the surveillance frequencies are relocated are still 
required to be operable, meet the acceptance criteria for the 
surveillance requirements, and be capable of performing any 
mitigation function assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased.
    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 previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The 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. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumption and current plant operating practice.
    Therefore, the proposed changes do 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 design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and Bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, NextEra 
Energy Duane Arnold will perform a probabilistic risk evaluation 
using the guidance contained in NRC approved NEI 04-10, Rev. 1 in 
accordance with the SFCP. NEI 04-10, Rev. 1, methodology provides 
reasonable acceptance guidelines and methods for evaluating the risk 
increase of proposed changes to surveillance frequencies consistent 
with Regulatory Guide 1.177.
    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: Ms. Marjan Mashhadi, 801 Pennsylvania 
Avenue, NW., Suite 220 Washington, DC 20004.
    NRC Branch Chief: Robert J. Pascarelli.

NextEra Energy Point Beach, LLC (the Licensee), Docket Nos. 50-266 and 
50-301, Point Beach

    Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks, Manitowac 
County, Wisconsin.
    Date of amendment request: March 23, 2011.
    Description of amendment request: The proposed amendment consists 
of replacing non-conservative values for five operating limits in the 
Technical Specifications with more conservative values that incorporate 
measurement uncertainty. Additionally, one of the operating limits will 
replace a volume expressed in cubic feet with a volume expressed in 
tank percent level to allow the plant operators a direct verification 
of the technical specification limit based on instrument readings.
    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 changes clarify the requirements for five plant 
operating limits by incorporating measurement uncertainties in the 
Technical Specification values to ensure the parameters remain 
within the ranges assumed in the accident analysis. The parameters 
are not accident initiators. Therefore, the proposed change will not 
increase the probability of an accident previously evaluated. 
Maintaining the parameters within the ranges specified in the 
Technical Specifications ensures that the systems will respond as 
assumed to mitigate the accidents previously evaluated. Therefore, 
the proposed change will not increase the consequences of an 
accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendment 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 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. The change does not alter assumptions made in the safety 
analysis, but ensures that plant operating parameters will be 
maintained as assumed in the accident analysis. The proposed change 
is consistent with the accident analysis assumptions.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment clarifies the requirements for plant 
operating limits by incorporating instrument uncertainties to ensure 
the parameters remain within the initial operating limits or ranges 
assumed in the accident analysis. No change is made to the accident 
analysis assumptions.

    Therefore, the proposed change would 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, Senior Attorney, NextEra 
Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: February 17, 2011, as supplemented on 
April 21, 2011.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7.1, ``Main Steam Safety Valves 
(MSSVs),'' Table 3.7.1-1, ``Maximum Allowable Power Range Neutron Flux 
High Setpoint With Inoperable MSSVs,'' and the Bases section for the 
MSSVs. This license amendment request proposed to remove a one-time 
note listed in TS Table 3.7.1-1, specific to Diablo Canyon Power Plant, 
Unit No. 2 for Cycle 15, that is no longer applicable or needed. This 
license amendment request also proposes to revise the TS Bases B 3.7.1 
to reflect a new analysis methodology for establishing the reduced 
Power Range Neutron Flux High setpoint for one inoperable MSSV as 
listed in TS Table 3.7.1-1. The supplement dated April 21, 2011, 
proposes to revise the Final Safety

[[Page 28476]]

Analysis Report Update (FSARU) Sections 15.2.7.3 and 15.2.16 to reflect 
the proposed changes to the TS Bases. The supplement provided 
additional information that clarified the application and did not 
expand the scope of the February 17, 2011, application.
    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?
    This License Amendment Request (LAR) proposes to remove a one-
time Unit 2 Cycle 15 Limiting Condition for Operation (LCO) 
exemption that is no longer applicable and revise the safety 
analysis performed in support of Technical Specification (TS) 3.7.1, 
``Main Steam Safety Valves (MSSVs),'' Table 3.7.1-1, ``Maximum 
Allowable Power Range Neutron Flux High Setpoint with Inoperable 
MSSVs'' for one inoperable MSSV. The revised safety analysis 
resolves a nonconforming condition associated with the TS 3.7.1 
Bases and re-establishes that the Power Range Neutron Flux High 
setpoint of 87 percent Rated Thermal Power (RTP) continues to 
provide adequate protection for one inoperable MSSV on each steam 
lead.
    The Power Range Neutron Flux High setpoint TS value does not 
initiate an accident. Technician adjustments to lower the Power 
Range Neutron Flux High setpoint could cause a reactor trip; 
however, this action is already a TS requirement. There has been no 
change in the TS setpoint value from the current value or in the 
requirement for a technician to adjust the setpoints downward when 
MSSVs become inoperable.
    Therefore, this proposed change will not increase the 
probability of a reactor trip.
    The revised TS B 3.7.1 safety analyses establishes that the 
current Power Range Neutron Flux High setpoint of 87 percent with 
one inoperable MSSV on each loop will ensure the remaining MSSVs 
will continue to prevent overpressure of the main steam leads and 
steam generators, and remove adequate heat from the RCS [reactor 
coolant system].
    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?
    The revised safety analysis which credits the Class 1 Over 
Temperature Delta Temperature (OTDT) reactor trip and the Power 
Range Neutron Flux High setpoint TS value with one inoperable MSSV 
do not initiate an accident and do not change the method by which 
any safety-related system performs its function.
    The proposed change does not result in plant operation outside 
the limits previously considered, nor allow the progression of 
transients or accidents in a manner different than previously 
considered.
    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 change involve a significant reduction in a margin 
of safety?
    The proposed change and revised safety analysis demonstrate that 
all applicable Reactor Coolant System (RCS) and steam generator (SG) 
pressure boundary acceptance criteria are satisfied, and re-
establish that the existing Power Range Neutron Flux High setpoint 
TS value for one inoperable MSSV remains conservatively bounding.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.
    With the proposed change, the MSSVs will prevent SG pressure 
from exceeding 110 percent of SG design pressure in accordance with 
the American Society of Mechanical Engineers code. The conclusions 
for the Final Safety Analysis Report accident analyses are 
unaffected by the change, remain valid, and provide margin.
    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, California 94120.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: March 14, 2011.
    Description of amendment request: The proposed amendments would 
revise the licenses and the Technical Specifications regarding Residual 
Heat Removal (RHR) and Coolant Circulation-Low Water Level, 
specifically, to allow one RHR loop to be inoperable for up to 2 hours 
for surveillance testing provided the other RHR loop is operable and in 
operation. The proposed change is described in Technical Specification 
Task Force Traveler TSTF-361-A, Revision 2, ``Allow standby SDC/RHR/DHR 
[shut down cooling/residual heat removal/decay heat removal] loop to 
[be] inoperable to support testing,'' approved for use by the Nuclear 
Regulatory Commission in a letter dated October 31, 2000 (Agencywide 
Documents Access and Management System, Accession No. ML003775261).
    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 adds an LCO [Limiting Condition for 
Operations] Note to LCO 3.9.6, ``RHR and Coolant Circulation-Low 
Water Level,'' to allow one RHR loop to be inoperable for up to 2 
hours for surveillance testing provided the other RHR loop is 
Operable and in operation. An inoperable RHR train is not an 
initiator to any accident previously evaluated. The RHR trains are 
not credited with mitigating any accident previously evaluated in 
Mode 6. As a result, the consequences of any accident previously 
evaluated are not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds an LCO Note to LCO 3.9.6, ``RHR and 
Coolant Circulation-Low Water Level,'' to allow one RHR loop to be 
inoperable for up to 2 hours for surveillance testing provided the 
other RHR loop is Operable and in operation. This allowance 
currently appears in Specification 3.4.7 and 3.4.8 and the 
conditions under which the Note would be applied in Specification 
3.9.6 are not significantly different from those specifications. The 
Note is needed in LCO 3.9.6 to provide the flexibility to perform 
surveillance testing while ensuring that there is reasonable time 
for operators to respond to and mitigate any expected failures.
    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

[[Page 28477]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Gloria Kulesa.

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

    Date of amendment request: July 12, 2010.
    Description of amendment request: The proposed revision is an 
administrative change that: (1) Corrects an error in TS 3.12.E.5, (2) 
deletes duplicative requirements in TS 3.12.E.2 and TS 3.12.E.4, (3) 
relocates the shutdown margin value in TS 3.12 and the TS 3.12 Basis to 
the Core Operating Limits Report (COLR), and (4) expands the TS 6.2 
list of parameters defined in the COLR.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The 
proposed LAR does not involve a physical change to any structures, 
systems, or components (SSCs) at Surry Power Station; nor does it 
change any of the previously evaluated accidents in the Updated 
Final Safety Analysis Report (UFSAR).
    Thus, this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed change is administrative in nature. The 
proposed change does not involve a physical change to any SSCs, and 
there is no impact on their design function. The proposed change 
does not affect initiators of analyzed events.
    Therefore, the proposed change does not introduce any new 
failures that could 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?
    No. The proposed change is administrative in nature. Margin of 
safety is established through the design of plant SSCs, the 
parameters within which the plant is operated, and the establishment 
of the setpoints for the actuation of equipment relied upon to 
respond to an event. The proposed change does not impact the 
condition or performance of SSCs relied upon for accident mitigation 
or any safety analysis assumptions.
    Therefore, the proposed amendment 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Gloria Kulesa.

Notice of Issuance of Amendments to Facility Operating 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.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, 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 records will be accessible from the 
Agencywide Documents Access and Management System (ADAMS) online at 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 Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].

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

    Date of application for amendment: April 13, 2010, as supplemented 
by a letter dated January 18, 2011.
    Brief description of amendment: The licensee proposed to revise 
section 3.1.a.1.C, ``Reactor Coolant Pumps,'' section 3.1.a.3, 
``Pressurizer Safety Valves,'' and section 3.1.b, ``Heatup and Normal 
Cooldown Limit Curves for Normal Operation,'' of the Technical 
Specifications (TS), as described in its application of April 13, 2010. 
After conversion of the TS to Improved Technical Specifications (ITS), 
the affected information was contained in ITS section 3.4.3, ``Reactor 
Coolant System (RCS) Pressure and Temperature (P-T, or equivalently P/
T) Limits'', ITS section 3.4.5, ``RCS Loops--MODE 3'', ITS section 
3.4.6, ``RCS Loops--MODE 4'', ITS section 3.4.10, ``Pressurizer Safety 
Valves'', ITS 3.4.12, ``Low Temperature Overpressure Protection (LTOP) 
System'', and ITS section 3.5.2, ``ECCS--Operating,'' as described in 
the licensee's supplement of January 18, 2011. Specifically, the 
proposed amendment would replace the heatup and cooldown pressure-
temperature (P-T) limit curves with new ones, and specify a higher LTOP 
enabling temperature. The supplement also provided additional 
restrictions on RCS mass addition until the reactor coolant system cold 
leg temperature exceeded 356 [deg]F, consistent with Improved Standard 
Technical Specifications.
    Date of issuance: April 29, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 208.
    Renewed Facility Operating License No. DPR-43: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 29, 2010 (75 FR 
37473). The supplement dated January 18, 2011, provided additional 
information that

[[Page 28478]]

clarified the application, did not expand the scope of the application, 
and did not change the Commission's proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 29, 2011.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2, Pope County, Arkansas

    Date of application for amendment: March 31, 2010, as supplemented 
by letters dated June 23, June 24, August 9, and September 16, 2010.
    Brief description of amendment: The amendment modified the 
requirements of the Technical Specification definitions, requirements, 
and terminology related to the use of an Alternate Source Term (AST) 
associated with accident offsite and control room dose consequences. In 
addition, implementation of the AST supports adoption of the control 
room envelope habitability controls in accordance with NRC-approved 
Technical Specification Task Force (TSTF)-448, Revision 3, ``Control 
Room Habitability.''
    Date of issuance: April 26, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 293.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: June 29, 2010 (75 FR 
37475). The supplemental letters dated June 23, June 24, August 9, and 
September 16, 2010, 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 April 26, 2011.
    No significant hazards consideration comments received: No.

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: November 8, 2010.
    Brief description of amendment: The amendment revised the Technical 
Specifications to be consistent with the NRC-approved Technical 
Specifications Task Force (TSTF) change traveler TSTF-493, ``Clarify 
Application of Setpoint Methodology for LSSS [Limiting Safety System 
Setting] Functions,'' Revision 4, Option A. Under Option A, two 
surveillance notes are added to TS Table 3.3.5.1-1, ``Emergency Core 
Cooling System Instrumentation,'' Function 3.d, ``Condensate Storage 
Tank Level--Low,'' and to TS Table 3.3.5.2-1, ``Reactor Core Isolation 
Cooling System Instrumentation,'' Function 3, ``Condensate Storage Tank 
Level--Low,'' for the suction swap from the condensate storage tank 
(CST) to the suppression pool function for the high pressure core spray 
and reactor core isolation cooling function, respectively. 
Specifically, surveillance notes would be added to surveillance 
requirements that require verifying trip setpoint setting values (i.e., 
channel calibration and trip unit calibration). The amendment completes 
a commitment made by the licensee to address an unresolved issue 
associated with TS Amendment No. 181 for the CST level-low setpoint 
change approved by the NRC in its letter dated February 25, 2009 (ADAMS 
Accession No. ML090290209).
    Date of issuance: April 27, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 185.
    Facility Operating License No. NPF-29: The amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 28, 2010 (75 
FR 81670).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2011.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2 (BVPS-2), Beaver County, 
Pennsylvania

    Date of application for amendment: April 9, 2009, as supplemented 
by letters dated June 15, 2009, January 18, 2010, March 18, 2010, May 
3, 2010, May 21, 2010, June 1, 2010, August 9, 2010, October 7, 2010, 
October 18, 2010, January 5, 2011, February 18, 2011, March 18, 2011, 
and March 21, 2011.
    Brief description of amendment: The amendment modified Technical 
Specifications (TSs) to support the replacement of existing Boraflex 
neutron absorber fuel storage racks in the BVPS-2 spent fuel pool with 
new high density, Metamic neutron absorber fuel storage racks, which 
will increase the total storage locations from 1,088 to 1,690.
    Date of issuance: April 29, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No: 173.
    Facility Operating License No. NPF-73: The amendment revised the 
License and the TSs.
    Date of initial notice in Federal Register: March 11, 2010 (75 FR 
11566). The supplements dated June 15, 2009, January 18, 2010, March 
18, 2010, May 3, 2010, May 21, 2010, June 1, 2010, August 9, 2010, 
October 7, 2010, October 18, 2010, January 5, 2011, February 18, 2011, 
March 18, 2011, and March 21, 2011, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 29, 2011.
    No significant hazards consideration comments received: No.

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 application for amendments: April 7, 2009, as supplemented 
by letters dated October 17, 2008; April 8, June 17 (2 letters), August 
24, September 11, September 25, October 9, November 13, November 20 (2 
letters), November 21 (2 letters), December 8, December 16, December 
21, and December 22 of 2009; January 7, January 8, January 13, January 
22, January 29, February 11, February 12, February 25, March 3, March 
24, March 25, April 15, April 21, April 22, April 26, April 28 (2 
letters), April 29, April 30, May 6, May 13, May 14, May 20, June 10 (2 
letters), June 11, June 14, June 24, July 8 (2 letters), July 15 (2 
letters), July 21, July 23, July 27, July 28, July 29, August 2, August 
6, August 9 (2 letters), August 12, August 23, August 24 (2 letters), 
August 26, September 1, September 8, September 9, September 14, 
September 21, September 27, September 28 (3 letters), October 1, 
October 12, October 14, October 15, October 28, November 1, November 4, 
November 12 (2 letters), November 15, November 30, December 1, December 
7, December 10 (2 letters), December 13, December 15, December

[[Page 28479]]

21 (2 letters), and December 30 of 2010; January 7 (2 letters), January 
11, January 13, January 21, February 22, March 2, and March 4 of 2011.
    Brief description of amendments: The proposed amendments would 
increase the licensed core power level for PBNP Units 1 and 2 from 1540 
megawatts thermal (MWt) to 1800 MWt. The increase in core thermal power 
will be approximately 17 percent over the current licensed thermal 
power level and is defined as an Extended Power Uprate (EPU). The 
proposed amendments would change the Renewed Facility Operating 
Licenses, the Technical Specifications (TSs) and licensing bases to 
support operation at the increased core thermal power level, including 
changes to the maximum licensed reactor core thermal power, reactor 
core safety limits, Constant Axial Offset Control (CAOC) operating 
strategy, Reactor Protection System (RPS) and Engineered Safety Feature 
Actuation System (ESFAS) Limited Safety System Settings (LSSSs) and 
diesel generator (DG) start loss of voltage time delays. Additional TS 
changes include Reactor Coolant System (RCS) flow rate, pressurizer 
operating level, pressurizer safety valve settings, accumulator and 
refueling water storage tank boron concentrations, main steam safety 
valve maximum allowable power level and lift settings, new Main 
Feedwater Isolation Valves (MFIVs), and Core Operating Limits Report 
(COLR) references.
    The review of the EPU LAR will include the changes to the HELB 
methodology to verify compliance with the licensing basis and 
acceptability for EPU conditions. The HELB evaluations have been re-
evaluated at EPU conditions using the following: (1) Implementation of 
NRC Generic Letter (GL) 87-11, ``Relaxation in Arbitrary Intermediate 
Pipe Rupture Requirements,'' dated June 19, 1987, and Branch Technical 
Position MEB 3-1, ``Postulated Rupture Locations in Fluid System Piping 
Inside and Outside Containment,'' Revision 2, dated June 1987, (2) mass 
and energy released from a HELB, (3) compartment pressurization 
transient evaluation following a HELB event, (4) jet impingement from 
streams following a HELB event, and (5) operator response time 
evaluation.
    Date of issuance: May 3, 2011.
    Effective date: Unit 1--As of the date of issuance and shall be 
implemented prior to Unit 1 startup from the Fall 2011 refueling 
outage. Unit 2--As of the date of issuance and shall be implemented 
prior to startup from the Spring 2011 refueling outage.
    Amendment Nos.: 241, 245.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revise the License, Appendix C, and the Technical 
Specifications.
    Date of initial notice in Federal Register: November 17, 2010 (75 
FR 70305).
    The supplemental letters contained clarifying information and did 
not change the staff's initial proposed finding of no significant 
hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 3, 2011.
    No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire

    Date of amendment request: June 28, 2010.
    Description of amendment request: This amendment revises the 
Seabrook Technical Specifications (TSs) by deleting TS 3/4.8.4.2, 
``Containment Penetration Conductor Overcurrent Protective Devices and 
Protective Devices for Class 1E Sources Connected to Non-Class 1E 
Circuits,'' and relocates the information to the Seabrook Technical 
Requirements Manual.
    Date of issuance: April 29, 2011.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 125.
    Facility Operating License No. NPF-86: The amendment revised the TS 
and the License.
    Date of initial notice in Federal Register: November 2, 2010 (75 FR 
67403).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 29, 2011.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: June 28, 2010.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.7.7, ``Control Room Makeup and Cleanup Filtration 
System,'' to add shutdown actions if the required actions for an 
inoperable control room envelope (CRE) boundary were not met. The 
amendments also added a note to the required action for an inoperable 
CRE boundary to clarify that the boundary is not a required system, 
subsystem, train, component, or device that depends on a diesel 
generator as a source of emergency power.
    Date of issuance: April 25, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1--195; Unit 2--183.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 21, 2010 (75 
FR 57529).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 2011.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 30, 2010, as supplemented 
by letters dated August 23, 2010, and March 4, 2011.
    Brief Description of amendments: The amendments revised the 
Technical Specifications by relocating specific surveillance frequency 
requirements to a licensee-controlled document using a risk-informed 
justification.
    Date of issuance: April 29, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: Unit 1--273 and Unit 2--272.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 10, 2010 (75 FR 
48377).
    The supplements dated August 23, 2010, and March 4, 2011, 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 April 29, 2011.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of May 2011.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2011-11804 Filed 5-16-11; 8:45 am]
BILLING CODE 7590-01-P