Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 68206-68224 [E7-23225]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 232 (Tuesday, December 4, 2007)]
[Notices]
[Pages 68206-68224]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-23225]


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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 November 8, 2007 to November 21, 2007. The 
last biweekly notice was published on November 20, 2007 (72 FR 65360).

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. Within 60 days after the date of publication of this 
notice, the licensee 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 for a hearing and a petition for leave to intervene.
    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 O1F21, 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 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

[[Page 68207]]

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 Viewer(tm) to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ 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 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

[[Page 68208]]

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.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina
    Date of amendments request: January 22, 2007, as supplemented by 
letters dated June 21, July 18, July 31, and October 15, 2007.
    Description of amendments request: The amendment would revise the 
Technical Specifications to support the transition to AREVA NP fuel and 
core design methodologies.
    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 amendments revise the list of NRC-approved 
analytical methods used to establish core operating limits. Core 
operating limits are established to ensure that fuel design limits 
are not exceeded during operating transients or accidents. The 
analytical methods used to determine core operating limits are those 
methods that have previously been found acceptable by the NRC and 
are required to be listed in the Technical Specification section 
governing the Core Operating Limits Report. The application of these 
NRC-approved analytical methods will continue to ensure that 
acceptable operating limits are established and applied to operation 
of the reactor core.
    The proposed amendments will add a new Technical Specification 
3.2.3, ``Linear Heat Generation Rate (LHGR),'' for fuel bundles, add 
a new definition to Technical Specification 1.1 for LHGR, and revise 
Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions 
on LHGR when in single recirculation loop operation or with an 
inoperable Turbine Bypass System. These LHGR limits will be 
established using NRC-approved analytical methods to ensure that 
fuel performance during normal, transient, and accident conditions 
is acceptable.
    Based on the above, the proposed amendments do not involve an 
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.
    As previously stated, the proposed amendments support transition 
from Global Nuclear Fuels Americas (GNF-A) fuel and core design and 
analysis services to AREVA NP fuel and core design and analysis 
services. The AREVA NP fuel assemblies which will be used in the 
BSEP Unit 1 and 2 cores will be similar in design to the GNF-A fuel 
that will be co-resident in the cores. The BSEP, Unit 1 and 2 cores 
in which this fuel will operate will be designed to meet all 
applicable design and licensing criteria. Adherence to these design 
and licensing criteria will not introduce any new modes of operation 
or introduce any new accident precursors, and thus will preclude the 
introduction 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 amendments will continue to require that core 
operating limits be determined using NRC-approved analytical 
methods. Acceptable fuel performance is obtained by ensuring that 
the peak cladding temperature (PCT) during a postulated design basis 
loss-of-coolant accident (LOCA) is maintained less than the limits 
specified in 10 CFR 50.46, and that the core remains in a coolable 
geometry following a postulated design basis LOCA. The proposed 
amendments ensure that adequate margin will continue to be 
maintained to the 2200 degree PCT limit of 10 CFR 50.46, and the use 
of NRC-approved analytical methods will continue to ensure 
acceptable fuel performance during normal operations, as well as 
during transient and accident conditions. Therefore, the proposed 
amendments 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 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: August 6, 2007.
    Description of amendments request: The amendment would revise the 
Technical Specifications (TSs) to implement Technical Specification 
Task Force (TSTF) Change TSTF-343, Revision 1, which allows the 
performance of visual examinations of the primary containment to be 
performed in accordance with the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 
Subsections IWE and IWL. The amendment would also make an 
administrative change to the TSs by eliminating a one-time requirement 
to perform containment leak rate testing that has already been 
completed.
    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 frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Primary Containment Leakage Rate Testing 
Program. In addition, the proposed change allows those examinations 
to be performed during power operation as opposed to during a 
refueling outage. The frequency of visual examinations of the 
metallic and concrete surfaces of the containment and the mode of 
operation during which those examinations are performed has no 
relationship to or adverse impact on the probability of any of the 
initiating events assumed in the accident analyses. The proposed 
change would allow

[[Page 68209]]

visual examinations that are performed in accordance with NRC-
approved ASME Section XI Code requirements, except where relief has 
been granted by the NRC, to meet the intent of visual examinations 
specified by Regulatory Guide 1.163, without requiring additional 
visual examinations in accordance with the Regulatory Guide. The 
intent of early detection of deterioration will continue to be met 
by the more vigorous requirements of the Code-required visual 
examinations. As such, the safety function of the containment as a 
fission product barrier is maintained.
    The proposed change also includes the removal of an item in TS 
5.5.12 which was incorporated to establish deadlines for performing 
the performance-based Type A leakage tests in conjunction with 
changing, on a one-time basis, the Type A test frequency. The 
specified Unit 1 and Unit 2 Type A test have been completed. As 
such, removal of this item is an administrative change.
    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. Therefore, 
based on the above, the proposed change does not represent a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the Primary Containment Leakage Rate 
Testing Program in TS 5.5.12 for consistency with the requirements 
of 10 CFR 50.55a(g)(4) for components classified as Code Class MC 
and CC. The proposed change affects the frequency of visual 
examinations that will be performed for the metallic and concrete 
surfaces of containment and allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage.
    The proposed change does not involve a modification to the 
physical configuration of the plants (i.e., no new equipment will be 
installed), and does not revise the methods governing normal plant 
operation. Also, the proposed change will not impose any new or 
different requirements or introduce a new accident initiator, 
accident precursor, or malfunction mechanism.
    The proposed change also includes the removal of an item in TS 
5.5.12 which was incorporated to establish deadlines for performing 
the performance based Type A leakage tests in conjunction with 
changing, on a one-time basis, the Type A test frequency. The 
specified Unit 1 and Unit 2 Type A test have been completed. As 
such, removal of this item is an administrative change.
    As such, 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 the Primary Containment Leakage Rate 
Testing Program in TS 5.5.12 for consistency with the requirements 
of 10 CFR 50.55a(g)(4) for components classified as Code Class MC 
and CC. The proposed change allows some of those examinations to be 
performed during power operation as opposed to during a refueling 
outage. As previously stated, the proposed change does not involve a 
modification to the physical configuration of the plants and does 
not revise the methods governing normal plant operation. As such, 
the safety function of the containment as a fission product barrier, 
will be maintained and is not adversely impacted by the proposed 
change.
    The proposed change also includes the removal of an item in TS 
5.5.12 which was incorporated to establish deadlines for performing 
the performance-based Type A leakage tests in conjunction with 
changing, on a one-time basis, the Type A test frequency. The 
specified Unit 1 and Unit 2 Type A test have been completed. As 
such, removal of this item is an administrative change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336 Millstone Power 
Station, Unit No. 2, New London County, Connecticut
    Date of amendment request: February 20, 2007.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit No. 2 (MPS2) Technical 
Specifications (TS) to eliminate Surveillance Requirement (SR) 4.5.2.e 
which requires flow rate verification for each charging pump. Charging 
pump flow is no longer relied upon for design basis mitigation at MPS2 
and the charging pumps have been classified as non-risk significant in 
the MPS2 Probabilistic Risk Assessment model. Therefore, the proposed 
amendment is requesting to remove the charging pump flow verification 
requirements currently located in the TS SR 4.5.2.e.
    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 FSAR [Final Safety Analysis Report] Chapter 14 accident 
analyses for MPS2 take no credit for the flow delivered by the 
charging pumps. Additionally, the proposed change does not modify 
any plant equipment or method of operation for any system, structure 
or component required for safe operation of the facility or 
mitigation of accidents assumed in the facility safety analyses. As 
such, the proposed amendment does not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not modify any plant equipment or 
method of operation for any system, structure or component required 
for safe operation of the facility or mitigation of accidents 
assumed in the facility safety analyses. As such, no new failure 
modes are introduced by the proposed change. Consequently, the 
proposed amendment does not introduce any accident initiators or 
malfunctions that would cause a new or different kind of accident. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The FSAR Chapter 14 accident analyses for MPS2 take no credit 
for the charging pumps. The TS change does not involve a significant 
reduction in a margin of safety because the proposed change does not 
affect equipment design or operation, and there are no changes being 
made to the technical specification required safety limits or safety 
system settings. The proposed change does not affect any of the 
assumptions used in the accident analysis, nor does it affect any 
method of operation for equipment important to plant safety. 
Therefore, the margin of safety is not impacted by the proposed 
amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 68210]]

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut.
    Date of amendment request: July 2, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) 4.0.5 to reference the American 
Society of Mechanical Engineers (ASME) Code for Operation and 
Maintenance of Nuclear Power Plants (OM Code) instead of Section XI of 
the ASME Boiler and Pressure Vessel Code. Specifically, the proposed 
amendment would modify the inservice inspection (ISI) of ASME Code 
Class 1, 2, and 3 components and inservice testing of ASME Code Class 
1, 2, and 3 pumps and valves to reflect the requirements in the ASME OM 
Code. In addition, the redundant requirement in TS 4.0.5 to maintain an 
ISI program is being proposed for removal, based on duplicate 
regulatory requirements set forth in Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.55a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not modify any plant equipment and does 
not impact any failure modes that could lead to an accident. 
Additionally, the proposed change has no effect on the consequence 
of any analyzed accident since the change does not affect the 
function of any equipment credited for accident mitigation. The 
proposed change incorporates revisions to the ASME Code that result 
in a net improvement in the measures for testing pumps and valves. 
Removing from TS the duplicate requirement in the regulations to 
maintain an ISI program in accordance with ASME codes and standards 
does not impact any accident initiators or analyzed events or 
mitigation of events. No reduction in previous commitments to 10 CFR 
50.55a(g) are being proposed by this change.
    Based on the discussion above, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or adversely affect 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. The proposed change does not 
alter existing test criteria or frequencies. Additionally, there is 
no change in the types or increases in the amounts of any effluent 
that may be released off-site and there is no increase in individual 
or cumulative occupational exposure. The proposed changes 
incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. Removal of 
the duplicate TS requirement to maintain an ISI program will not 
alter the commitment to the current ISI program requirements in 10 
CFR 50.55a or any other TS requirements related to inservice 
inspection.
    Based on the discussion above, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises TS 4.0.5 regarding inservice testing 
of ASME Code Class 1, 2, and 3 pumps and valves, for consistency 
with the requirements of 10 CFR 50.55a(f)(4). The proposed change 
incorporates an administrative clarification to the frequencies for 
IST and incorporates revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. No 
setpoints or safety limit settings are being revised. The safety 
function of the affected pumps and valves will continue to be 
confirmed through inspection and testing. Removal of the ISI program 
requirement from TS 4.0.5 does not remove the requirement from 
regulations, and therefore, will not diminish the current station 
approved programs and procedures that implement the regulatory 
criteria of 10 CFR 50.55a(g) to maintain an acceptable ISI program 
in accordance with the ASME Code.
    Based on the discussion above, 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: Lillian M. Cuoco, Esquire, Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Building 475, 5th Floor, 
Rope Ferry Road, Waterford, CT 06141-5127.
    NRC Branch Chief: Harold K. Chernoff.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Oconee Nuclear Station Independent Spent Fuel Storage Installation NRC 
License No. SNM-2503, Docket No. 72-4, Oconee County, South Carolina
    Date of amendment request: March 14, 2007.
    Description of amendment request: The amendments would revise the 
licenses to reflect the change in the name of the licensee from Duke 
Power Company LLC to Duke Energy Carolinas, LLC. The proposed 
amendments are a name change only. There is no change in the state of 
incorporation, registered agent, registered office, rights or 
liabilities of the company. Nor is there a change in the function of 
the licensee or the way in which it does business.
    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 amendments are for a name change only. The 
amendments do not involve any change in the technical qualifications 
of the licensee or the design, configuration, or operation of the 
nuclear units. All Limiting Conditions for Operation, Limiting 
Safety System Settings and Safety Limits specified in the Technical 
Specifications remain unchanged. Also, the Physical Security Plans 
and related plans, the Operator Training and Requalification 
Programs, the Quality Assurance Programs, and the Emergency Plans 
will not be materially changed by the proposed name change. The name 
change amendments will not affect the executive oversight provided 
by the Chief Nuclear Officer and his staff.
    Therefore, the proposed amendments do not involve any increase 
in the probability or consequences of an accident previously 
analyzed.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.

[[Page 68211]]

    The proposed amendments do not involve any change in the design, 
configuration, or operation of the nuclear plant. The current plant 
design, design bases, and plant safety analysis will remain the 
same.
    The Limiting Conditions for Operations, Limiting Safety System 
Settings and Safety Limits specified in the Technical Specifications 
are not affected by the proposed changes. As such, the plant 
conditions for which the design basis accident analyses were 
performed remain valid.
    The proposed amendments do not introduce a new mode of plant 
operation or new accident precursors, do not involve any physical 
alterations to plant configurations, or make changes to system 
setpoints that could initiate a new or different kind of accident.
    Therefore, the proposed amendments 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 amendments do not involve a change in the design, 
configuration, or operation of the nuclear plants. The change does 
not affect either the way in which the plant structures, systems, 
and components perform their safety function or their design and 
licensing bases.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications. Because there is no 
change to the physical design of the plant, there is no change to 
any of these margins.
    Therefore, the proposed amendments do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: November 7, 2007.
    Description of amendment request: The proposed amendment would 
delete License Condition 2.F, which requires reporting of violations of 
certain other requirements contained in Section 2.C of the license.
    The NRC staff issued a ``Notice of Availability of Model 
Application Concerning Elimination of Typical License Condition 
Requiring Reporting of Violations of Section 2.C of Operating License 
Using the Consolidated Line Item Improvement Process'' in the Federal 
Register on November 4, 2005 (70 FR 67202). The notice referenced a 
model safety evaluation, a model no significant hazards consideration 
(NSHC) determination, and a model license amendment request published 
in the Federal Register on August 29, 2005 (70 FR 51098). In its 
application dated November 7, 2007, 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 adopted by the licensee 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 involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase 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 is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the 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 deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2 (IP2), Westchester County, New York
    Date of amendment request: October 24, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements related to the 
containment buffering agent used for pH control under post loss-of-
coolant accident (LOCA) conditions. Specifically, the proposal would 
approve the use of sodium tetraborate (STB) as the buffering agent 
instead of the currently approved compound, trisodium phosphate (TSP). 
The reason for this change in buffering agents is to minimize the 
potential for an adverse chemical interaction between the TSP and 
certain insulation materials in the containment that could degrade flow 
through the sump screens following certain design-basis accident 
scenarios such as a LOCA.
    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 amendment does not involve a significant increase 
in the probability of an accident previously evaluated because the 
containment buffering agent is not an initiator of any analyzed 
accident. The proposed change does not impact any failure modes that 
could lead to an accident.
    The proposed amendment does not involve a significant increase 
in the consequences of an accident previously evaluated. The 
buffering agent in containment is designed to buffer the acids 
expected to be produced after a LOCA and is credited in the 
radiological analysis for iodine retention. Utilizing STB as a 
buffering agent ensures the post LOCA containment sump mixture will 
have a pH >= 7.0. The proposed change of replacing TSP with STB 
results in the radiological consequences remaining within the limits 
of 10 CFR 50.67 as demonstrated by existing analyses of record.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response--No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. STB is a passive component that is proposed to be used at 
IP2 as a buffering agent to increase the pH of the initially acidic 
post-LOCA containment water to a more neutral pH. Changing the 
proposed buffering agent from TSP to STB does not constitute an 
accident initiator or create a new or different

[[Page 68212]]

kind of accident previously analyzed. The proposed amendment does 
not involve operation of any required systems, structures or 
components in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the changes being requested.
    Therefore, the proposed amendment 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 amendment does not involve a significant reduction 
in a margin of safety. The proposed amendment of changing the 
buffering agent from TSP to STB results in equivalent control of 
maintaining sump pH at 7.0 or greater, thereby controlling 
containment atmosphere iodine and ensuring the radiological 
consequences of a LOCA are within regulatory limits. The use of STB 
also reduces the potential for exacerbating sump screen blockage due 
to a chemical interaction between TSP and certain calcium sources 
used in containment. This proposed amendment eliminates the 
formation of calcium phosphate precipitate thereby reducing the 
overall amount of precipitate that may be formed in a postulated 
LOCA. The buffer change would minimize the potential chemical 
effects and should enhance the ability of the emergency core cooling 
system to perform the post-accident mitigating functions.
    Therefore, the proposed amendment does not involve a significant 
reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York
    Date of amendment request: October 24, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements regarding the setpoint 
and definition of the low-low level alarm on the Refueling Water 
Storage Tank (RWST). Specifically, the proposal would revise the 
setpoint of the low-low level alarm from a range of greater than or 
equal to 10.5 ft and less than or equal to 12.5 ft to a range of 
greater than or equal to 9.0 ft and less than or equal to 11.0 ft, and 
revise the definition of the RWST ``low level alarm'' to ``low-low 
level alarm.'' The reason for these changes is to ensure that adequate 
water is supplied to the containment floor to eliminate the risk of 
vortexing and/or draw down at the sump strainer modules following a 
small-break loss-of-coolant accident (LOCA). The proposed changes are 
being requested to support resolution of the pressurized-water reactor 
sump performance issue involving debris accumulation, Generic Safety 
Issue (GSI)-191.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications are 
consistent with the assumptions of all design basis accidents, as 
they exist currently and as affected subsequent to the 
implementation of the proposed amendment. The change in the RWST 
low-low level alarm setpoint range has been demonstrated to be 
within the safety margins for post-accident parameters and, in most 
cases, actually beneficial to plant post-accident response 
capability. The RWST is designed to respond to a variety of 
accidents, and, for operation in Modes 1 through 4, it serves no 
other purpose. Therefore, any adjustment of an intermediate level 
setpoint cannot increase the probability of a design basis accident. 
The change in the definition of the RWST ``low level alarm'' to 
``low-low level alarm'' is editorial and therefore does not affect 
the function of the alarm. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes represent a minor adjustment to an existing 
setpoint range. The effect of the changes will be to assure 
recirculation flow following a LOCA with consideration for sump 
strainer installation, in response to GSI-191. However, the RWST 
will continue to perform its function in essentially the same manner 
that it has since original plant design. No changes in equipment 
operation or procedural control will result from this amendment that 
could possibly degrade the performance of the RWST or cause it to be 
operated in a manner inconsistent with existing design basis 
assumptions. The change in the definition of the RWST ``low level 
alarm'' to ``low-low level alarm'' is editorial and therefore does 
not affect the function of the alarm. Therefore, the proposed 
changes do 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 changes improve the margin to safety, especially 
with respect to post-accident temperature/pressure and dose 
consequences during injection and, most importantly, pump 
performance under postulated sump debris conditions during 
recirculation. Significant margin is available to preclude air 
ingestion in the ECCS [emergency core cooling system] pumps, and 
sufficient time is available for the operators to perform the 
switchover to recirculation. The change in the definition of the 
RWST ``low level alarm'' to ``low-low level alarm'' is editorial and 
therefore does not affect the function of the alarm. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania


[[Page 68213]]


Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: July 19, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Sections 5.3.1/6.3.1, ``Unit Staff 
Qualifications,'' for operator license applicants in accordance with 
current industry standards for education and experience eligibility 
requirements. The proposed amendment would permit changes to the unit 
staff qualification education and experience eligibility requirements 
for licensed operators. The proposal will bring Exelon Generation 
Company, LLC (EGC) and AmerGen Energy Company, LLC (AmerGen) sites in 
alignment with current industry practices and facilitate the 
development of a pre-initial licensed operator training program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    Licensed operator qualification and training can have an 
indirect impact on accidents previously evaluated. However, the NRC 
considered this impact during the rulemaking process, and by 
promulgation of the revised 10 CFR 55 rule, determined that this 
impact remains acceptable when licensees have an accredited licensed 
operator training program which is based on a systems approach to 
training (SAT). The NRC has concluded in RIS [Regulatory Issue 
Summary] 2001-01 and NUREG-1021 that standards and guidelines 
applied by INPO [the Institute of Nuclear Power Operations] in their 
accredited training programs are equivalent to those put forth by or 
endorsed by the NRC. Therefore, maintaining an INPO accredited SAT 
licensed operator training program is equivalent to maintaining an 
NRC approved licensed operator training program which conforms with 
applicable NRC Regulatory Guidelines or NRC endorsed industry 
standards. The proposed changes conform to ACAD [air containment 
atmosphere distribution] 00-003, Revision 1 licensed operator 
education and experience eligibility requirements.
    Based on the above, EGC and AmerGen conclude that the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the licensed operator 
training programs, which are administrative in nature. The EGC and 
AmerGen licensed operator training programs have been accredited by 
INPO and are based on SAT.
    Based on the above discussion, EGC and AmerGen conclude that the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed TS changes are administrative in nature. The 
proposed TS changes do not affect plant design, hardware, system 
operation, or procedures for accident mitigation systems. The 
proposed changes do not impact the performance or proficiency 
requirements for licensed operators. As a result, the ability of the 
plant to respond to and mitigate accidents is unchanged by the 
proposed TS changes. Therefore, these changes do not involve a 
significant reduction in a margin of safety.
    Based on the above, EGC and AmerGen conclude that the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    Based on the above evaluation of the three criteria, EGC and 
AmerGen conclude that the proposed amendment presents no significant 
hazards consideration under the standards set forth in 10 CFR 
50.92(c), and, accordingly, a finding of ``no significant hazards 
consideration'' is justified.

    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353, 
Limerick Generating Station, Unit 1 and 2, Montgomery County, 
Pennsylvania

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,York 
and Lancaster Counties, Pennsylvania

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: August 8, 2007.
    Description of amendment request: The proposed amendment replaces 
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) in the applicable technical specification (TS) section for the 
Inservice Testing Program (IST) for the Exelon Generation Company, LLC, 
and AmerGen Energy Company, LLC, (the licensees) plants that have 
implemented industry Improved Technical Specifications. The proposed 
changes are based on TS Task Force (TSTF) 479-A, Revision 0, ``Changes 
to Reflect Revision of 10 CFR 50.55a,'' as modified by TSTF-497, 
Revision 0, ``Limit Inservice Testing Program SR [Surveillance 
Requirement] 3.0.2 Application to Frequencies of 2 Years or Less.'' In 
addition, the proposed amendment adds a provision in the applicable TS 
section to only apply the extension allowance of SR 3.0.2 to the 
frequency table listed in the TS as part of the IST and to normal and 
accelerated inservice testing frequencies of two years or less, as 
applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the applicable TS Section to conform 
to the requirements of 10 CFR 50.55a, ``Codes and

[[Page 68214]]

standards,'' paragraph (f) regarding the inservice testing of pumps 
and valves. The current TS reference the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes 
would reference the ASME OM Code as applicable, which is consistent 
with 10 CFR 50 .55a, paragraph (f), ``Inservice testing 
requirements.'' In addition, the proposed changes clarify that the 
extension allowance of SR 3.0.2 only applies to the frequency table 
listed in the TS, if applicable, as part of the Inservice Testing 
Program and to normal and accelerated inservice testing frequencies 
of two years or less. The definitions of the frequencies are not 
changed by this license amendment request.
    The proposed changes are administrative in nature, do not affect 
any accident initiators, do not affect the ability to successfully 
respond to previously evaluated accidents and do not affect 
radiological assumptions used in the evaluations. Thus, the 
probability or radiological consequences of any accident previously 
evaluated are not increased.
    Therefore, the proposed changes do 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 changes revise the applicable TS Section to conform 
to the requirements of 10 CFR 50.55a(f) regarding the inservice 
testing of pumps and valves. The current TS Section references the 
ASME Boiler and Pressure Vessel Code, Section XI, requirements for 
the inservice testing of ASME Code Class 1, 2, and 3 pumps and 
valves. The proposed changes would reference the ASME OM Code as 
applicable, which is consistent with 10 CFR 50.55a(f). In addition, 
the proposed changes clarify that the extension allowance of SR 
3.0.2 only applies to the frequency table listed in the TS, if 
applicable, as part of the Inservice Testing Program and to normal 
and accelerated inservice testing frequencies of two years or less. 
The definitions of the frequencies are not changed by this license 
amendment request.
    The proposed changes to the applicable TS Section do not affect 
the performance of any structure, system, or component credited with 
mitigating any accident previously evaluated and do not introduce 
any new modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes revise the applicable TS Section for 
Braidwood Station Units 1 and 2, Byron Station Units 1 and 2, 
Dresden Nuclear Power Station Units 2 and 3, Limerick Generating 
Station Units 1 and 2, Oyster Creek Generating Station, Peach Bottom 
Atomic Power Station Units 2 and 3, Quad Cities Nuclear Power 
Station Units 1 and 2, and Three Mile Island Unit 1 to conform to 
the requirements of 10 CFR 50.55a(f) regarding the inse