Biweekly Notice; Applications and Amendments to Facility Operating Licenses; Involving No Significant Hazards Considerations, 58669-58684 [E8-23342]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 195 (Tuesday, October 7, 2008)]
[Notices]
[Pages 58669-58684]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-23342]


=======================================================================
-----------------------------------------------------------------------

 NUCLEAR REGULATORY COMMISSION


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 September 11, 2008 to September 24, 2008. 
The last biweekly notice was published on September 23, 2008 (73 FR 
54862).

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, 
Directives and Editing Branch, 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 delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
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 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. 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

[[Page 58670]]

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 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.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated on August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve 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 technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    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

[[Page 58671]]

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). To 
be timely, filings must be submitted no later than 11:59 p.m. Eastern 
Time on the due date.
    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. 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) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: August 28, 2008.
    Description of amendments request: The amendment would relocate the 
main steam isolation valve times in Technical Specification (TS) 
section 3.7.2, ``Main Steam Isolation Valves (MSIVs)'' to the licensee 
controlled document that is referenced in the Bases. In addition, the 
valve isolation times in the TS are replaced with the phrase ``within 
limits.'' The changes are consistent with the Nuclear Regulatory 
Commission approved Technical Specification Task Force (TSTF)-491, 
Revision 2, ``Removal of Main Steam and Main Feedwater Valve Isolation 
Times From Technical Specifications.'' The availability of the TS 
improvement was published in the Federal Register on December 29, 2006 
(71 FR 250) as part of the consolidated item improvement process 
(CLIIP).
    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 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 phrase, ``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, tests 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 phrase ``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 pant 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 phrase ``within limits.'' Instituting 
the proposed changes will continue to ensure the testing of main 
steam and main feedwater isolation valves. Changes to the Bases are 
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 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel-Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: June 19, 2008.
    Description of amendments request: The proposed change would: (1) 
Revise Technical Specifications (TS) control rod notch surveillance 
requirement (SR) frequency in TS 3.1.3, ``Control Rod Operability,'' 
and (2) revise Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify 
the applicability of the 1.25 surveillance test extension. The licensee

[[Page 58672]]

is proposing to adopt the approved Technical Specification Task Force 
(TSTF) change traveler TSTF-475, Revision 1, ``Control Rod Notch 
Testing Frequency.'' A notice of availability of TSTF-475, Revision 1, 
was published in the Federal Register on November 13, 2007 (72 FR 
63935).
    In addition, the proposed amendment would remove Note 2 associated 
with SR 3.1.3.3 for Unit 1, which is a cycle-specific note and has 
expired. This change is administrative in nature and does not affect 
the no significant hazards consideration (NSHC) determination.
    Basis for proposed NSHC determination: As required by 10 CFR 
50.91(a), the licensee, in its application dated June 19, 2008, 
affirmed the applicability of the published model NSHC determination, 
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 generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and 
NUREG-1434 (BWR/6) STS. The changes: (1) revise TS testing frequency 
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY'', (2) [not applicable to BSEP], and (3) revise Example 
1.4-3 in Section 1.4 ``Frequency'' to clarify the applicability of 
the 1.25 surveillance test interval extension. The consequences of 
an accident after adopting TSTF-475, Revision 1 are no different 
than the consequences of an accident prior to adoption. Therefore, 
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 does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously analyzed. Thus, 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
    TSTF-475, Revision 1 will: (1) revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY'', (2) [not 
applicable to BSEP], and (3) revise Example 1.4-3 in Section 1.4 
``Frequency'' to clarify the applicability of the 1.25 surveillance 
test interval extension. The GE Nuclear Energy Report, ``CRD 
Notching Surveillance Testing for Limerick Generating Station,'' 
dated November 2006, concludes that extending the control rod notch 
test interval from weekly to monthly is not expected to impact the 
reliability of the scram system and that the analysis supports the 
decision to change the surveillance frequency. Therefore, the 
proposed changes in TSTF-475, Revision 1 [. . .] do not involve a 
significant reduction in a margin of safety.

    Based on the review of the above analysis, the NRC staff finds that 
the three standards in 10 CFR 50.92(c) are satisfied. Therefore, the 
NRC staff proposes to determine that the amendment request involves 
NSHC.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II-Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: June 19, 2008.
    Description of amendments request: The proposed change would revise 
Limiting Condition for Operation (LCO) 3.10.1, and the associated 
Bases, to expand its scope to include provisions for temperature 
excursions greater than 212 degrees Fahrenheit ([deg]F) as a 
consequence of inservice leak and hydrostatic testing, and as a 
consequence of scram time testing initiated in conjunction with an 
inservice leak or hydrostatic test, while considering operational 
conditions to be in Mode 4.
    The NRC issued a ``Notice of Availability of Model Application on 
Technical Specification Improvement to Modify Requirements Regarding 
LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation Using the 
Consolidated Line Item Improvement Process,'' associated with Technical 
Specification Task Force (TSTF) Improved Standard Technical 
Specification Change Traveler, TSTF-484, Revision 0, in the Federal 
Register on October 27, 2006 (71 FR 63050). The NRC also issued a 
Federal Register notice on August 21, 2006 (71 FR 48561) that provided 
a model safety evaluation and a model no significant hazards 
consideration (NSHC) determination relating to modification of 
requirements regarding LCO 3.10.1, ``Inservice Leak and Hydrostatic 
Testing Operation.'' In its application dated June 19, 2008, the 
licensee affirmed the applicability of the model NSHC determination.
    Basis for proposed NSHC determination: As required by 10 CFR Part 
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
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or consequences of an accident 
previously evaluated. 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
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. 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 or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. 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
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    Based on the above, the proposed change presents NSHCs under the 
standards set forth in 10 CFR 50.92(c). Therefore, the NRC staff 
proposes to determine that the amendment request involves NSHC.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II-

[[Page 58673]]

Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.

Entergy Nuclear Operations, Inc., Docket No. 50-3, Indian Point Energy 
Center, Unit 1 Westchester County, New York

    Date of amendment request: June 26, 2008.
    Description of amendment request: The proposed amendment would 
delete license conditions and Technical Specification (TS) requirements 
which relate to the storage of spent nuclear fuel in the Indian Point 
Unit 1 (IP1) Fuel Handling Building Spent Fuel Pool. The spent fuel is 
to be transferred to, and stored at, the existing Indian Point 
Independent Spent Fuel Storage Installation (ISFSI), Docket No. 72-51. 
The removal of the stored spent fuel and drain down of the spent fuel 
pools renders many of the license conditions and TS requirements 
unnecessary and burdensome.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The proposed changes are all contingent on the prior removal 
of the stored spent fuel from the IP1 Spent Fuel Pool (SFP) to the 
Indian Point Energy Center (IPEC) ISFSI. The accidents previously 
evaluated in the IP1 Final Safety Analysis Report (FSAR), which 
consists of the IP1 Decommission Plan and Supplemental Environmental 
Information, are stored fuel related accidents. The removal of the 
stored fuel from the IP1 facility to the IPEC ISFSI precludes the 
possibility of these accidents.
    Consequently, the proposed changes to the license do not involve 
a significant increase in the probability or the consequences of an 
accident previously evaluated.
    2. Does the proposed change create the probability of a new or 
different accident from any accident previously evaluated?
    The proposed changes are all contingent on the prior removal of 
the stored spent fuel from the IP1 SFP to the IPEC ISFSI. With the 
removal of the stored spent fuel from the IP1 facility, and 
considering the IP1 has been in a SAFESTOR mode for over thirty 
years, no significant source term remains which could result in any 
postulated radiological event that would impact the health and 
safety of the public.
    Therefore, the proposed changes to the IP1 license consequently 
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 a 
margin of safety?
    No. The proposed changes are all contingent on the prior removal 
of the stored spent fuel from the IP1 SFP to the IPEC ISFSI. Upon 
the removal of spent fuel, the Technical Specifications being 
deleted no longer are required to protect the health and safety of 
the public or occupational workers from the potential adverse 
conditions, hazards or accidents as discussed in the FSAR.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant reduction in the 
margin of safety.

    Based upon the reasoning presented above 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: Assistant General Counsel, Entergy Nuclear 
Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
    NRC Branch Chief: Theodore Smith, Acting.

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 amendment request: September 11, 2008.
    Description of amendment request: The amendment would revise 
several surveillance requirements (SRs) and add SR 3.8.1.21 in 
Technical Specification (TS) 3.8.1, ``AC [alternating current] 
Sources--Operating,'' and TS 3.8.2, ``AC Sources--Shutdown.'' The 
amendment would allow the slow-start testing sequence of the diesel 
generators in order to reduce the stress and wear on the equipment.
    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 affects the surveillance requirements for 
the Diesel Generators (DGs). The DGs are onsite standby power 
sources intended to provide redundant and reliable power to ESF 
[Engineered Safety Feature] systems credited as accident mitigating 
features in design basis [accident] analyses. Per NRC Regulatory 
Guide (RG) 1.9, Revision 3, which is referenced in Grand Gulf 
Nuclear Station (GGNS) UFSAR [Updated Final Safety Analysis Report 
Section] 8.3.1.2.1, the proposed change is intended to allow slower 
starts of the DGs during testing in order to reduce DG aging effects 
due to excessive testing conditions. As such, the proposed change 
will result in improved DG reliability and availability, thereby 
providing additional assurance that the DGs will be capable of 
performing their safety function. The method of starting the 
emergency diesel generators for testing purposes does not affect the 
probability of any previously evaluated accident. Although the 
change allows slower starts for the monthly tests, the more rapid 
start function, assumed in the accident analysis, is unchanged and 
will be verified on a 184 day frequency. Therefore the accident 
analysis consequences are not affected [by the proposed change].
    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 affects the surveillance requirements for 
the onsite ac sources, i.e. the Diesel Generators. Accordingly, the 
proposed change does not involve any change to the configuration or 
method of operation of any plant equipment that could cause an 
accident. In addition, no new failure modes have been created nor 
has any new limiting failure been introduced as a result of the 
proposed surveillance changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change is intended to bring the existing GGNS TS 
requirements for the onsite AC sources in line with regulatory 
guidance. Under the proposed change, the DGs will remain capable of 
performing their safety function, and the effects of aging on the 
DGs will be reduced by eliminating unnecessary testing. The DG start 
times assumed in the current accident analyses are unchanged and 
will be verified on a 184 day frequency.
    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: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340

[[Page 58674]]

Echelon Parkway, Jackson, Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

GE Hitachi Nuclear Energy (GEH), License No. DR-10, Docket No. 50-183, 
ESADA Vallecitos Experimental Superheat Reactor (EVESR)

    Date of amendment request: June 23, 2008.
    Description of amendment request: The proposed license amendment 
would modify the Technical Specification (TS) requirements to revise 
the scope of dismantling activities that GEH can perform under The 
Vallecitos Nuclear Center Liabilities Reduction Project and specify 
radiological control requirements of 10 CFR Part 20. Two TS changes are 
proposed. The proposed changes to the TS:
     Allow GEH to conduct dismantling activities below the 549-
ft elevation level within the containment building; and
     Revise the physical security requirements for access to 
areas below the 549-ft elevation level within the containment building.
    The application for license amendment is available electronically 
at the NRC's Electronic Reading Room at https://www.nrc.gov/reading-rm/
adams.html. From this site, you can access the NRC's Agencywide 
Document Access and Management System (ADAMS), which provides text and 
image files of NRC's public documents. The ADAMS accession number for 
the June 23, 2008, request is ML081780099.
    If you do not have access to ADAMS, or if there are problems in 
accessing the documents located in ADAMS, contact the NRC Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737, or 
by e-mail to pdr@nrc.gov. These documents may also be viewed 
electronically on the public computers located at the NRC's PDR, 01F21, 
One White Flint North, 11555 Rockville Pike, Rockville, MD 20852. The 
PDR reproduction contractor will copy documents for a fee.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Proposed change one is an administrative change submitted to 
clarify the area where dismantling activities will occur as 
authorized by the facility license. The majority of the component 
removal activities will occur in areas below the 549-ft. elevation. 
Proposed change two removes the specific shielding and covering 
requirements for the reactor vessel, shield plug storage pit and the 
empty spent fuel storage pit and modifies the access control 
requirements to be consistent with 10 CFR 20. The EVESR reactor was 
shutdown in 1967 and has remained in a ``Possess Only'' status. All 
fuel bundles were removed from the facility and the radiation and 
contamination levels have been reduced by the removal of radioactive 
material and natural decay. No aspect of the proposed changes will 
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 evaluated?
    Response: No.
    Proposed change one is an administrative, therefore there it 
cannot create a new or different kind of accident. Removal of the 
specific shielding and covering requirements for the reactor vessel, 
shield plug storage pit and the empty spent fuel storage pit and 
modification of the access control requirements as described in 
proposed change two will not impact the function or integrity of the 
reactor pressure vessel, which is the primary safety system required 
to be maintained by the license. The proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident evaluated.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    Removal of the specific shielding, covering and access control 
requirements will not result in a reduction of the margin of the 
safety for the EVESR facility. These controls were implemented to 
provide shielding and access controls to High Radiation Areas. Since 
the reactor is no longer operating and the radiological conditions 
have been significantly reduced, the specific controls specified in 
the current technical specifications are not required. All areas in 
the EVESR containment will be controlled in accordance with 10 CFR 
20. High Radiation areas will be controlled in a manner consistent 
with the requirements of 10 CFR 20.1601. The proposed changes do not 
affect the margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based upon 
the staff's review of the licensee's analysis, as well as the staff's 
own evaluation, the staff concludes 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.
    GEH, Manager, Regulatory Compliance & EHS: LaTonya L. Mahlahla.
    NRC Branch Chief: Andrew Persinko.

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: June 24, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by (1) replacing the 
references to Section XI of the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code with references to the 
ASME Code for Operation and Maintenance of Nuclear Power Plants (OM 
Code); and (2) revising the allowance to extend Inservice Testing (IST) 
frequencies by 25 percent to clearly state that the allowance is 
applicable to IST frequencies of 2 years or less. The proposed changes 
are based on TS Task Force (TSTF) Standard Technical Specification 
Change Traveler 479-A, Revision 0, ``Limit Inservice Testing Program SR 
3.0.2 Application to Frequencies of 2 Years or Less.''
    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 revise the IST Program sections of the NMP1 
and NMP2 TS to maintain consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the IST of pumps and valves that are 
classified as ASME Code Class 1, Class 2, and Class 3. The proposed 
changes incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. The 
proposed changes also revise the allowance to extend IST frequencies 
by 25 percent to clearly state that this allowance is applicable to 
IST frequencies of 2 years or less.
    The proposed TS changes are administrative in nature. They do 
not impact any accident initiators, the ability to mitigate 
previously evaluated accidents, or the assumptions used in 
evaluating the radiological consequences of previously evaluated 
accidents. The proposed changes do not involve the addition or 
removal of any equipment, or any design changes to the facilities.
    Therefore, the proposed changes do 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.

[[Page 58675]]

    The proposed changes revise the IST Program sections of the NMP1 
and NMP2 TS to maintain consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the IST of pumps and valves that are 
classified as ASME Code Class 1, Class 2, and Class 3. The proposed 
changes incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. The 
proposed changes also revise the allowance to extend IST frequencies 
by 25 percent to clearly state that this allowance is applicable to 
IST frequencies of 2 years or less.
    The proposed TS changes are administrative in nature. They do 
not involve a modification to the physical configuration of the 
plants (i.e., no new equipment will be installed) or involve a 
change in the methods governing normal plant operation. The proposed 
changes will not impose any new or different requirements or 
introduce a new accident initiator, accident precursor, or failure 
mechanism.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS changes are administrative in nature. They do 
not involve a modification to the physical configuration of the 
plants (i.e., no new equipment will be installed) or change the 
methods governing normal plant operation. The proposed changes do 
not modify the safety limits or setpoints at which protective 
actions are initiated, and do not change the requirements governing 
operation or availability of safety equipment assumed to operate to 
preserve margins of safety. The incorporation of revisions to the 
ASME Code results in a net improvement in the measures for testing 
pumps and valves. The safety function of the affected pumps and 
valves will be maintained.
    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: 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: July 7, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) testing frequency for the 
Surveillance Requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' 
The proposed change revises the frequency of SR 3.1.4.2, control rod 
scram time testing, from ``120 days cumulative operation in Mode 1'' to 
``200 days cumulative operation in Mode 1.'' These changes are based on 
TS Task Force (TSTF) change traveler TSTF-460 (Revision 0) that has 
been approved generically for the Boiling-water reactor (BWR) Standard 
TS, NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency 
of SR 3.1.4.2, control rod scram time testing, from ``120 days 
cumulative operation in MODE 1'' to ``200 days cumulative operation in 
MODE 1.'' The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination (NSHCD) for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864) using the consolidated line item improvement 
process (CLIIP). The licensee affirmed the applicability of the model 
NSHC determination and the model safety evaluation in its application 
dated July 7, 2008.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC, based on the model NSHCD published in the Federal Register on 
August 23, 2004 (69 FR 51864), is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident, previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation.
    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 continues to test the control rod scram time 
to ensure the assumptions in the safety analysis are protected. 
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.

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: July 7, 2008.
    Description of amendment request: PPL Susquehanna, LLC (the 
licensee) requests adoption of the Nuclear Regulatory Commission (NRC) 
approved Technical Specification Task Force (TSTF) change traveler 
TSTF-475, (Revision 1), ``Control Rod Notch Testing Frequency and SRM 
[Source Range Monitor] Insert Control Rod Action,'' to change the 
Standard Technical Specifications (STS) for General Electric (GE) 
Plants (NUREG-1433, BWR/4 to the plant specific TS, that allows: (1) 
Revising the frequency of Surveillance Requirement (SR) 3.1.3.2, notch 
testing of fully withdrawn control rod, from ``7 days after the control 
rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM'' to 
``31 days after the control rod is withdrawn and THERMAL POWER is 
greater than the LPSP [Low Power Set Point] of the RWM [Rod With 
Minimizer]'', and (2) revising Example 1.4-3 in Section 1.4 
``Frequency'' to clarify that the 1.25 surveillance test interval 
extension in SR 3.0.2 is applicable to time periods discussed in NOTES 
in the ``SURVEILLANCE'' column in addition to the time periods in the 
``FREQUENCY'' column.
    The NRC staff issued a notice of availability in the Federal 
Register on November 13, 2007, (72 FR 63935), which included a model 
safety evaluation (SE) and model no significant hazards consideration 
determination (NSHCD), using the consolidated line-item improvement 
process (CLIIP), of possible amendments to revise the plant specific 
TS, to allow: (1) Revising the frequency of SR 3.1.3.2, notch testing 
of fully withdrawn control rod, from ``7 days after the control rod

[[Page 58676]]

is withdrawn and THERMAL POWER is greater than the LPSP of RWM'' to 
``31 days after the control rod is withdrawn and THERMAL POWER is 
greater than the LPSP of the RWM'', (2) adding the word ``fully'' to 
LCO 3.3.1.2 Required Action E.2 to clarify the requirement to fully 
insert all insertable control rods in core cells containing one or more 
fuel assemblies when the associated SRM instrument is inoperable, and 
(3) revising Example 1.4-3 in Section 1.4 ``Frequency'' to clarify that 
the 1.25 surveillance test interval extension in SR 3.0.2 is applicable 
to time periods discussed in NOTES in the ``SURVEILLANCE'' column in 
addition to the time periods in the ``FREQUENCY'' column. The licensee 
affirmed the applicability of the model SE and model NSHC determination 
in its application dated July 7, 2008.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC, based on the model NSHCD published in the Federal Register on 
November 13, 2007 (72 FR 63935), 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 generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and NUREG-
1434 (BWR/6) STS. The changes: (1) revise TS testing frequency for 
surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY'', (2) clarify the requirement to fully insert all 
insertable control rods for the limiting condition for operation (LCO) 
in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring 
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. Implementing TSTF-475, Revision 1 
does not change the control rod notch test method. Implementing TSTF-
475, Revision 1 decreases the performance frequency of the control rod 
notch test. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated. The consequences of an accident after adopting TSTF-475, 
Revision 1 are no different than the consequences of an accident prior 
to adoption. Therefore, 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 does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The proposed 
change will not introduce new failure modes or effects and will not, in 
the absence of other unrelated failures, lead to an accident whose 
consequences exceed the consequences of accidents previously analyzed. 
Thus, 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 amendment will: (1) Revise the TS SR 3.1.3.2 frequency 
in TS 3.1.3, ``Control Rod OPERABILITY'', and (2) revise Example 1.4-3 
in Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. The GE Nuclear Energy Report, 
``CRD Notching Surveillance Testing for Limerick Generating Station,'' 
dated November 2006, concludes that extending the control rod notch 
test interval from weekly to monthly is not expected to impact the 
reliability of the scram system and that the analysis supports the 
decision to change the surveillance frequency. Therefore, the proposed 
changes in TSTF-475, Revision 1 do 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.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: July 21, 2008.
    Description of amendment request: The proposed amendments would 
delete the requirements related to plant staff working hours from 
Section 6.0, ``Administrative Controls'' of the respective plants' 
Technical Specifications (TSs). The current working hour requirements 
were incorporated into the TSs as a result of the guidance in Nuclear 
Regulatory Commission (NRC) Generic Letter (GL) 82-12, ``Nuclear Power 
Plant Staff Working Hours.'' The guidance in GL 82-12 has been 
superseded by the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), Part 26, ``Fitness for Duty Programs,'' Subpart 
I, ``Managing Fatigue'' which was published in the Federal Register on 
March 31, 2008, as part of the final rulemaking for Part 26. As 
discussed in the Federal Register notice for the final rule (73 FR 
16966), Subpart I must be implemented by licensees no later than 
October 1, 2009. The licensee stated that the proposed amendments would 
support implementation of the new requirements in Subpart I.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The removal of GL 82-12 administrative controls will not remove 
the requirement to control work hours and manage fatigue. Removal of 
TS controls required by GL 82-12 will be performed concurrently with 
the implementation of the more conservative [10 CFR Part 26], 
Subpart I, requirements. The proposed changes do 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. The proposed changes do 
not impact the initiators or assumptions of analyzed events, nor do 
they impact the mitigation of accidents or transient events.
    Because these new requirements are more conservative with 
respect to work hour controls and fatigue management, this will not 
significantly increase the probability or consequence of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes remove GL 82-12 administrative controls 
from [the] TS to support the implementation of Subpart I to [10 CFR 
Part 26]. The Subpart I regulations are more restrictive than the 
current guidance in [the] TS and would add conservatism to work hour 
controls and fatigue management. Work hours will continue to be 
controlled in accordance with NRC requirements. The new rule 
continues to allow for deviations from controls to mitigate or 
prevent a condition adverse to safety or necessary to maintain the 
security of the facility. This ensures that the new rule will not 
restrict work hours at the

[[Page 58677]]

expense of the health and safety of the public as well as plant 
personnel. The proposed changes do not alter plant configuration, 
require that new plant equipment be installed, alter assumptions 
made about accidents previously evaluated, add any initiators, or 
impact the function of plant SSCs or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected.
    Because the proposed changes do not remove the station's 
requirement to control work hours and increases the conservatism of 
work hour controls by changing administrative scheduling 
requirements, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    An input to maintaining the margin of safety is the control of 
work hours in managing fatigue. Salem and Hope Creek Generating 
Stations will continue their fitness-for-duty and behavioral 
observation programs, both of which will be strengthened by 
compliance with the new Part 26 regulation. The proposed changes add 
conservatism to fatigue management and contribute to the margin of 
safety. The proposed changes do not involve any physical changes to 
plant SSCs or the manner in which SSCs are operated, maintained, 
modified, tested, or inspected. The proposed changes do not involve 
a change to any safety limits, limiting safety system settings, 
limiting conditions of operation, or design parameters for any SSC. 
The proposed changes do not impact any safety analysis assumptions 
and do not involve a change in initial conditions, system response 
times, or other parameters affecting an accident analysis. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, with changes in the areas noted above, 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: July 30, 2008.
    Description of amendment request: The proposed amendment would 
relocate Technical Specification (TS) 3/4.7.5, ``Snubbers,'' to the 
Hope Creek Generating Station (HCGS) Technical Requirements Manual 
(TRM). TS 6.10.3.l, which specifies retention requirements for records 
of snubber service life monitoring pursuant to TS 4.7.5, would also be 
relocated to the TRM. In addition, the amendment would add new TS 
Limiting Condition for Operation (LCO) 3.0.8, ``Inoperability of 
Snubbers,'' and would modify LCO 3.0.1 to reference LCO 3.0.8.
    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 evaluate