Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 18251-18262 [E9-8832]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 75 (Tuesday, April 21, 2009)]
[Notices]
[Pages 18251-18262]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-8832]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0170]


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 staff) 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 March 26, 2009, to April 8, 2009. The last 
biweekly notice was published on April 7, 2009 (74 FR 15765).

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 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.
    Written comments may be submitted by mail to the Chief, Rulemaking 
and Directives Branch, 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. Copies of written comments 
received may be examined at the Commission's Public Document Room 
(PDR), located at One White Flint North, Public File Area O1F21, 11555 
Rockville Pike (first floor), Rockville, Maryland.
    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, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
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, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
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 
petitioner/requestor 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 petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor 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 petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one

[[Page 18252]]

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, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, 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, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). 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 a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital 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/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE 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 
EIE 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 may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday, excluding government holidays. The help 
electronic filing Help Desk can be contacted by telephone at 1-866-672-
7640 or by e-mail at MSHD.Resource@nrc.gov.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, 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.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii).
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a 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.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://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-

[[Page 18253]]

4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: August 21, 2008.
    Description of amendment request: The proposed amendments would 
revise the proposed license amendment implements Technical 
Specification Task Force (TSTF) Changes Travelers TSTF-479, Revision 0, 
``Changes to Reflect Revision of [Title 10 of the Code of Federal 
Regulations] 10 CFR 50.55a'' and TSTF-497, Revision 0, ``Limit 
Inservice Testing [IST] Program SR 3.0.2 Application to Frequencies of 
2 Years or Less''. TSTF-479 and TSTF-497 revise the technical 
specification Administrative Controls section pertaining to 
requirements for the IST Program, consistent with the requirements of 
10 CFR 50.55a(f)(4) for pumps and valves which are classified as 
American Society of Mechanical Engineers (ASME) Code Class 1, Class 2, 
and Class 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS [Technical Specification] 5.5.8, 
``Inservice Testing Program,'' for consistency with the requirements 
of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and 
valves which are classified as ASME Code Class 1, Class 2, and Class 
3. The proposed change incorporates revisions to the ASME [American 
Society of Mechanical Engineers] Code as identified in the TSTFs 
[Technical Specification Task Force] referenced above.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. The proposed change does not involve the addition or removal 
of any equipment, or any design changes to the facility. 
Additionally, there is no change in the types or increases in the 
amounts of any effluent that may be released offsite and there is no 
increase in individual or cumulative occupational exposure.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, Class 2, and Class 3. The 
proposed change incorporates revisions to the ASME Code as 
identified in the TSTFs referenced above. The proposed change does 
not involve a modification to the physical configuration of the 
plant nor does it involve a change in the methods governing normal 
plant operation. The proposed change will not impose any new or 
different requirements or introduce a new accident initiator, 
accident precursor, or malfunction mechanism. Additionally, there is 
no change in the types or increases in the amounts of any effluent 
that may be released offsite and there is no increase in individual 
or cumulative occupational exposure.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, Class 2, and Class 3. The 
proposed change does not involve a modification to the physical 
configuration of the plant nor does it change the methods 
governingnormal plant operation. The proposed change incorporates 
revisions to the ASME Code as identified in the TSTFs referenced 
above.
    The safety function of the affected pumps and valves will be 
maintained.
    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: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie Wong.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: February 16, 2009.
    Description of amendment request: The Arkansas Nuclear One, Unit 
No. 1 (ANO-1) Technical Specification (TS) 5.5.16, ``Reactor Building 
Leakage Rate Testing Program,'' contains reactor building leak rate 
criteria for overall Type A, B, and C testing. However, TS 5.5.16 does 
not specify criteria for Type B air lock leakage testing. Entergy 
Operations, Inc., proposes to modify TS 5.5.16 to add criteria for 
overall air lock leakage testing and to adopt a low pressure test 
method relevant to the air lock door seals. This change is consistent 
with NUREG 1430, Revision 3.1, ``Standard Technical Specifications 
(STS) for Babcock & Wilcox Plants.''
    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 reactor building air locks are passive components integral 
to the reactor building structure and are not associated with 
accident initiators. Each air lock door is rated for and tested to 
the maximum calculated post-accident pressure of the reactor 
building. The air lock door seal pressure test is performed any time 
the air lock is used for reactor building access during modes of 
operation when reactor building integrity is required and prior to 
establishing reactor building integrity. The door seal test is 
intended to be a gross test to verify that the door seals were not 
damaged during the opening and closing cycle(s). This test does not 
replace the required overall barrel leakage test. Based on 
information provided by the air lock vendor, a test pressure of 10 
psig [pounds per square inch gauge] is conservatively sufficient to 
perform this gross seal verification. This new acceptable leakage 
rate and test criteria are consistent with NUREG 1430, Rev. 3.1, 
Standard Technical Specifications for Babcock & Wilcox Plants (STS) 
and are applicable to ANO-1. While new to the TSs, the ANO-1 program 
for ensuring compliance with 10 CFR 50, Appendix J has verified 
leakage within the proposed limiting values.
    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 physical changes to the facility are initiated by the 
proposed change. In addition, the proposed change has no affect on 
plant configuration, or method of operation of plant structures, 
systems, or components.
    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.
    The proposed change does not increase the allowable overall air 
lock leakage rate, nor

[[Page 18254]]

affect the acceptance criteria of the overall integrated containment 
leakage rate as currently tested to in accordance with the ANO-1 
containment leakage rate test program. All of the changes are 
bounded by existing analyses for all evaluated accidents and do not 
create any situations that alter the assumptions used in these 
analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 10, 2009.
    Description of amendment request: The proposed amendment consists 
of changes to Technical Specification (TS) 3.4.9, ``Pressurizer,'' 
which contains a maximum and minimum level for the pressurizer. The 
licensee proposes to delete the minimum level requirement. This change 
is consistent with NUREG 1430, Rev. 3.1, ``Standard Technical 
Specifications [STS] for Babcock and Wilcox Plants.''
    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 minimum Pressure level limit currently specified in the TSs 
does not act to ensure specified fuel design limits are protected. 
Accident and transient analyses assume lowering or a loss of 
Pressurizer level. Safety systems are designed and maintained 
available to mitigate the consequences of an accident or transient 
that may involve a loss of Reactor Coolant System (RCS) inventory. 
None of these systems rely upon a predetermined minimum Pressurizer 
level in order to perform their intended function. Furthermore, the 
minimum Pressure level limit is unrelated to any anticipated 
accident initiator.
    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 physical changes to the facility are initiated by the 
proposed change. In addition, the proposed change has no affect on 
plant configuration, or method of operation of plant structures, 
systems, or components.
    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.
    Installed automatic control systems will continue to maintain 
Pressurizer level at a predetermined setpoint and are independent of 
a prescribed minimum TS level limit. The deletion of the current TS 
limit has no impact on guidance or operational response to 
pressurizer level deviations. Furthermore, the minimum Pressure 
level limit is not an assumed value for accident prevention or 
mitigation in the [Arkansas Nuclear One, Unit 1] [Safety Analysis 
Report].
    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: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: February 25, 2009.
    Description of amendment request: The proposed change removes the 
reactor coolant system (RCS) structural integrity requirements 
contained in Technical Specification (TS) 3/4.4.8, which specifies 
requirements relating to the structural integrity of American Society 
of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components. This 
specification is redundant to the requirements contained within Title 
10 of the Code of Federal Regulations (10 CFR) section 50.55a, ``Codes 
and standards.'' With this proposed change, RCS pressure boundary 
structural integrity will continue to be maintained by compliance with 
10 CFR 50.55a, as implemented through the Limerick Generating Station, 
Units 1 and 2, Inservice Inspection 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, with NRC edits in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to remove the RCS structural integrity 
controls from the TSs does not impact any mitigation equipment or 
the ability of the RCS pressure boundary to fulfill any required 
safety function. Since no accident mitigation [equipment] or 
initiators are impacted by this change, no design basis accidents 
are 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change will not alter the plant configuration or 
change the manner in which the plant is operated. No new failure 
modes are being introduced by the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Removal of TS 3/4.4.8 from the TSs does not reduce the controls 
that are required to maintain the RCS pressure boundary for ASME 
Code Class 1, 2, or 3 components.
    No equipment or RCS safety margins are impacted due to the 
proposed change. 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 18255]]

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: February 16, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by removing the structural 
integrity requirements contained in TS 3/4.4.10 and the associated TS 
bases from the TSs. Removal of TS 3/4.4.10 is consistent with NUREG-
1431, Revision 3.0, ``Standard Technical Specifications Westinghouse 
Plants,'' in that it does not meet the criteria of Title 10 of the Code 
of Federal Regulations (10 CFR), Part 50, Section 50.36, ``Technical 
Specifications,'' for inclusion in the TSs. The proposed amendment 
would also relocate the reactor coolant pump (RCP) flywheel inspection 
requirements in Surveillance Requirement (SR) 4.4.10 to SR 4.0.5, and 
would revise the RCP flywheel inspection interval from 10 years to 20 
years. The RCP flywheel inspection interval change is consistent with 
Nuclear Regulatory Commission approved Industry/Technical Specification 
Task Force (TSTF) Standard Technical Specification Change Traveler, 
TSTF-421, ``Revision to RCP Flywheel Inspection Program (WCAP-15666).''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to remove structural integrity controls from 
the TSs does not impact any mitigation equipment or the ability of 
the RCS [reactor coolant system] pressure boundary to fulfill any 
required safety function. The proposed change will continue to 
ensure the requirements of 10 CFR 50.55a [``Codes and standards''] 
are maintained as specified in TS 4.0.5. Since no accident 
mitigation or initiators are impacted by this change, no design 
basis accidents are 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 previously evaluated?
    Response: No.
    The proposed change will not alter the plant configuration or 
change the manner in which the plant is operated. Structural 
integrity will continue to be maintained as required by 10 CFR 
50.55a and specified in TS 4.0.5. No new failure modes are being 
introduced by the 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 
the margin of safety?
    Response: No.
    Removal of TS 3/4.4.10 from the TSs does not reduce the controls 
that are required to maintain the structural integrity of ASME 
[American Society of Mechanical Engineers] Code Class 1, 2, or 3 
components. No safety margins are impacted due to the proposed 
change.
    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 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 Branch Chief: Thomas H. Boyce.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket Nos. 50-220 and 
50-410, Nine Mile Point Nuclear Station Unit Nos. 1 and 2 (NMP 1 and 
2), Oswego County, New York

    Date of amendment request: February 11, 2009.
    Description of amendment request: The proposed amendment would 
delete those portions of the Technical Specifications (TSs) superseded 
by 10 CFR Part 26, Subpart I. The proposed change is consistent with 
Nuclear Regulatory Commission (NRC)-approved Revision 0 to TS Task 
Force (TSTF) Change Traveler, TSTF-511-A, ``Eliminate Working Hour 
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.'' 
The availability of the TS improvement was announced in the Federal 
Register (FR) on December 30, 2008 (73 FR 79923) as part of the 
consolidated line item improvement process. The licensee concluded that 
the no significant hazards consideration determination as presented in 
the FR notice is applicable to NMP 1 and 2.
    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:

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

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Removal of the 
Technical Specification requirements will be performed concurrently 
with the implementation of the 10 CFR Part 26, Subpart I, 
requirements. The proposed change does not impact the physical 
configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which the SSCs are operated, 
maintained, modified, tested, and inspected. Worker fatigue is not 
an initiator of any accident previously evaluated. Worker fatigue is 
not an assumption in the consequence mitigation 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.

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

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Working hours will 
continue to be controlled in accordance with NRC requirements. The 
new rule allows for deviations from controls to mitigate or prevent 
a condition adverse to safety or as necessary to maintain the 
security of the facility. This ensures that the new rule will not 
unnecessarily restrict working hours and thereby create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed change does not alter the plant 
configuration, require new plant equipment to be installed, alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

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

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Working hours will 
continue to be controlled in accordance with NRC requirements. The 
proposed change does not involve any physical changes to the plants 
or alter the manner in which plant systems are operated, maintained, 
modified, tested, and inspected. The proposed change does not alter 
the manner in which safety limits, limiting safety

[[Page 18256]]

system settings or limiting conditions for operation are determined. 
The safety analysis acceptance criteria are not affected by this 
change. The proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shutdown the plants 
and to maintain the plants in a safe shutdown condition. Removal of 
plant-specific Technical Specification administrative requirements 
will not reduce a margin of safety because the requirements in 10 
CFR Part 26 are adequate to ensure that worker fatigue is managed. 
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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: February 20, 2009.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements related to control 
room envelope habitability in TS 3.7.3, ``Plant Systems Control Room 
Emergency Outside Air Supply (CREOAS) System,'' and TS Section 5.5, 
``Administrative Controls Programs and Manuals.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on October 17, 2006 (71 FR 61075), on possible 
amendments to revise the plant specific TS, to strengthen TS 
requirements regarding control room envelope (CRE) habitability by 
changing the action and surveillance requirements associated with the 
limiting condition for operation operability requirements for the CRE 
emergency ventilation system. A new TS administrative controls program 
on CRE habitability is being added, including a model safety evaluation 
and model no significant hazards consideration determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on January 17, 
2007 (72 FR 2022). The licensee affirmed the applicability of the model 
NSHC determination in its application dated February 20, 2009.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented 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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. 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 significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the 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. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief : Mark Kowal.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: March 23, 2009.
    Description of amendment request: The proposed amendment would 
delete those portions of the Technical Specifications (TSs) superseded 
by Part 26, Subpart I of Title 10 of the Code of Federal Regulations 
(10 CFR). This change incorporates NRC approved Revision 0 of Technical 
Specification Task Force (TSTF) Improved Standard Technical 
Specification Change Traveler, TSTF-511, ``Eliminate Working Hour 
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.'' 
The availability of this TS improvement was announced as part of the 
consolidated line item improvement process (CLIIP) in the Federal 
Register on December 30, 2008 (73 FR 79923).
    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:

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

    The proposed change removes technical specification restrictions 
on working hours for personnel who perform safety related functions. 
The technical specification

[[Page 18257]]

restrictions are superseded by the worker fatigue requirements in 10 
CFR Part 26. Removal of the technical specification requirements 
will be performed concurrently with the implementation of 10 CFR 
Part 26, Subpart I, requirements. The proposed change does not 
impact the physical configuration or function of plant structures, 
systems, or components (SSCs) or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected. Worker fatigue 
is not an assumption in the consequence mitigation 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.

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

    The proposed change removes technical specification restrictions 
on working hours for personnel who perform safety related functions. 
The technical specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Working hours will 
continue to be controlled in accordance with NRC requirements. The 
new rule allows for deviations from controls to mitigate or prevent 
a condition adverse to safety or as necessary to maintain the 
security of the facility. This ensures that the new rule will not 
unnecessarily restrict working hours and thereby create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or effect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected.
    Therefore, the proposed change does 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 a Margin of Safety

    The proposed change removes technical specification restrictions 
on working hours for personnel who perform safety related functions. 
The technical specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26.
    The proposed change does not involve any physical changes to the 
plant or alter the manner in which plant systems are operated, 
maintained, modified, tested, or inspected. The proposed change does 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions or operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shutdown the plant 
and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific technical specification administrative 
requirements will not reduce a margin of safety because the 
requirements in 10 CFR Part 26 are adequate to ensure that worker 
fatigue is managed.
    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: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, 
Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: December 29, 2008.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.8.4, ``DC [Direct Current] Sources--
Operating,'' and TS 3.8.5, ``DC Sources--Shutdown.'' Specifically, this 
amendment would revise the battery connection resistance limits in 
Surveillance Requirement (SR) 3.8.4.2 and SR 3.8.4.5 from 150 micro-
ohms (150E-6 ohm) to 69 micro-ohms (69E-6 ohm). TS 3.8.5 is affected by 
virtue of SR 3.8.5.1 invoking both SR 3.8.4.2 and SR 3.8.4.5 for DC 
sources that are required to be operable in Modes 5 and 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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces a battery surveillance limit with a 
value based on voltage drop calculations for each of the four 
battery subsystems at Callaway under both normal operating and 
accident load profiles. The new value is more conservative, as well 
as being more appropriate, as an acceptance criterion for verifying 
battery operability pursuant to SR 3.8.4.2 and SR 3.8.4.5, thus 
providing greater assurance that the batteries can perform their 
specified safety functions with regard to accident mitigation.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no design changes. All design, material, and construction standards 
that were applicable prior to this amendment request will be 
maintained. There will be no changes to any design or operating 
limits.
    The proposed change will not adversely affect accident 
initiators or precursors, nor adversely alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is operated and maintained. The proposed 
change will not alter or prevent the ability of structures, systems, 
and components (SSCs) from performing their intended functions to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits.
    The proposed change does not physically alter safety-related 
systems nor affect the way in which safety-related systems perform 
their functions.
    All accident analysis acceptance criteria will continue to be 
met with the proposed change. The proposed change will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of any 
accident previously evaluated. The applicable radiological dose 
criteria will continue to be met.
    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.
    There are no proposed design changes nor are there any changes 
in the method by which any safety-related plant structure, system, 
or component (SSC) performs its specified safety function. The 
proposed changes will not affect the normal method of plant 
operation or change any operating parameters. Equipment performance 
necessary to fulfill safety analysis missions will be unaffected. 
The proposed change will not alter any assumptions required to meet 
the safety analysis acceptance criteria.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, or Solid State Protection System used in the plant 
protection systems.
    The proposed change does not, therefore, create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change does not involve a significant 
reduction in a margin of safety?
    Response: No.
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions. There will be no 
impact on the overpower limit, departure from nucleate boiling ratio 
(DNBR) limits, heat flux hot channel factor (FQ), nuclear 
enthalpy rise hot channel factor (F[Delta]H), loss of coolant 
accident peak cladding temperature (LOCA PCT), peak local power 
density, or any other margin of safety. The applicable radiological 
dose consequence acceptance criteria will continue to be met.

[[Page 18258]]

    The proposed change does not eliminate any surveillances or 
alter the frequency of surveillances required by the Technical 
Specifications; however, the acceptance criterion for the specified 
battery resistance surveillances will be more restrictive. None of 
the acceptance criteria for any accident analysis will be changed.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: March 4, 2009.
    Description of amendment request: The proposed amendment consists 
of changes to the approved fire protection program as described in Wolf 
Creek Generating Station (WCGS) Updated Safety Analysis Report (USAR). 
Specifically, a deviation from certain technical requirements of Title 
10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix R, 
Section III.G.2, as documented in Appendix 9.5E of the WCGS USAR, is 
requested regarding the use of operator manual actions in lieu of 
meeting circuit separation protection criteria. Table 3-1 of the 
submittal dated March 4, 2009 (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML090771269), identifies the 
proposed feasible and reliable operator manual actions requested for 
permanent approval and Table 3-2 of the submittal identifies the 
proposed feasible operator manual actions requested for approval on an 
interim basis. The interim operator actions will be eliminated with the 
implementation of associated design change package.
    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 design function of structures, systems and components are 
not impacted by the proposed change. The proposed change involves 
the performance of operator manual actions to achieve and maintain 
safe shutdown in the event of a fire outside of the control room and 
will not initiate an event. The proposed change does not increase 
the probability of occurrence of a fire or any other accident 
previously evaluated.
    The proposed operator manual actions are feasible and reliable 
and demonstrate that the plant can be safely shutdown in the event 
of a fire. No significant consequences result from the performance 
of the proposed 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The design function of structures, systems and components are 
not impacted by the proposed change. The proposed change involves 
the performance of operator manual actions to achieve and maintain 
safe shutdown in response to a fire outside of the control room. The 
operator manual actions do not involve new failure mechanisms or 
malfunctions that can initiate a new accident.
    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.
    For the permanent operator manual actions, adequate time is 
available to perform the proposed operator manual actions to account 
for uncertainties in estimates of the time available and in 
estimates of how long it takes to diagnose and execute the actions. 
The actions have been verified that they can be performed through 
demonstration and the actions are proceduralized. The proposed 
actions are feasible and reliable and demonstrate that the plant can 
be safely shutdown in the event of a fire.
    For the interim operator manual actions adequate time is 
available to feasibly perform the proposed operator manual actions 
and a compensatory measure fire watch is provided for the affected 
area as an added defense in depth fire protection measure.
    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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: March 6, 2009.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.2.2.d regarding the requirement 
to develop and implement administrative procedures to limit the working 
hours of personnel who perform safety-related functions. The 
requirements of TS 5.2.2 have been superseded by Title 10 of the Code 
of Federal Regulations (10 CFR) Part 26, Subpart I. The change is 
consistent with U.S. Nuclear Regulatory Commission (NRC)-approved 
Revision 0 to Technical Specification Task Force (TSTF) Improved 
Technical Specification Change Traveler, TSTF-511, ``Eliminate Working 
Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 
26.''
    The NRC staff issued a ``Notice of Availability of Model Safety 
Evaluation, Model No Significant Hazards Determination, and Model 
Application for Licensees That Wish To Adopt TSTF-511, Revision 0, 
`Eliminate Working Hour Restrictions From TS 5.2.2 To Support 
Compliance With 10 CFR Part 26,' '' in the Federal Register on December 
30, 2008 (73 FR 79923). The notice included a model safety evaluation, 
a model no significant hazards consideration (NSHC) determination, and 
a model license amendment request, using the consolidated line item 
improvement process. In its application dated March 6, 2009, the 
licensee affirmed the applicability of the model NSHC determination, 
which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC determination is presented 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 removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Removal of the 
Technical Specification requirements will be performed concurrently 
with the implementation of the 10 CFR Part

[[Page 18259]]

26, Subpart I, requirements. The proposed change does not impact the 
physical configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. Worker fatigue is not an 
initiator of any accident previously evaluated. Worker fatigue is 
not an assumption in the consequence mitigation 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.

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

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Working hours will 
continue to be controlled in accordance with NRC requirements. The 
new rule allows for deviations from controls to mitigate or prevent 
a condition adverse to safety or as necessary to maintain the 
security of the facility. This ensures that the new rule will not 
unnecessarily restrict working hours and thereby create the 
possibility of a new or different kind of accident from a