[Federal Register Volume 73, Number 88 (Tuesday, May 6, 2008)]
[Notices]
[Pages 25034-25050]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-9679]


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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 April 10 to April 23, 2008. The last 
biweekly notice was published on April 22, 2008 (73 FR 21567).

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

[[Page 25035]]

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 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one 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 in 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 
[email protected], 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

[[Page 25036]]

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 http://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 
http://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 http://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 http://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 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 
http://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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Arizona Public Service Company (APS), et al., Docket Nos. STN 50-528, 
STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of amendment request: January 17, 2008, as supplemented 
February 29, 2008.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TS) to establish more effective 
and appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task 
Force (TSTF) Standard Technical Specification change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' Specifically, the proposed 
amendments would modify TS 3.7.11, ``Control Room Essential Filtration 
System (CREFS),'' and add new TS 5.5.17, ``Control Room Envelope 
Habitability Program,'' to TS Administrative Controls Section 5.5, 
``Programs and Manuals.''
    The NRC staff issued a ``Notice of Availability of Technical 
Specification Improvement to Modify Requirements Regarding Control Room 
Envelope Habitability Using the Consolidated Line Item Improvement 
Process,'' associated with TSTF-448, Revision 3, in the Federal 
Register on January 17, 2007 (72 FR 2022). The notice included a model 
safety evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated January 17, 2008, 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:
Criterion 1--The Proposed Change[s] [Do] Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change[s] [do] not adversely affect accident 
initiators or precursors nor alter the design assumptions, conditions, 
or configuration of the facility. The proposed change[s] [do] not alter 
or prevent the ability of structures, systems, and components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the assumed

[[Page 25037]]

acceptance limits. The proposed change[s] [revise] the TS for the CRE 
[essential filtration] system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE [essential 
filtration] system is the CRE boundary. The CRE [essential filtration] 
system is not an initiator or precursor to any accident previously 
evaluated. Therefore, the probability of any accident previously 
evaluated is not increased. Performing tests to verify the operability 
of the CRE boundary and implementing a program to assess and maintain 
CRE habitability ensure that the CRE [essential filtration] system is 
capable of adequately mitigating radiological consequences to CRE 
occupants during accident conditions, and that the CRE [essential 
filtration] system will perform as assumed in the consequence analyses 
of design basis accidents. Thus, the consequences of any accident 
previously evaluated are not increased. Therefore, the proposed 
change[s] [do] not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change[s] [Do] Not Create the Possibility of 
a New or Different Kind of Accident From any Accident Previously 
Evaluated
    The proposed change[s] [do] not impact the accident analysis. The 
proposed change[s] [do] not alter the required mitigation capability of 
the CRE [essential filtration] system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new surveillance 
or following the new program. The proposed change[s] [do] not involve a 
physical alteration of the plant (i.e., no new or different type of 
equipment will be installed) or a significant change in the methods 
governing normal plant operation. The proposed change[s] [do] not alter 
any safety analysis assumptions and is consistent with current plant 
operating practice. Therefore, [the] change[s] [do] not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
Criterion 3--The Proposed Change[s] [Do] Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change[s] [do] not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change[s] [do] not affect safety 
analysis acceptance criteria. The proposed change[s] will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The proposed 
change[s] [do] not adversely affect systems that respond to safely shut 
down the plant and to maintain the plant in a safe shutdown condition. 
Therefore, the proposed change[s] [do] not involve a significant 
reduction in a margin of safety. Based upon the reasoning presented 
above and the previous discussion of the amendment request, the 
requested change does not involve a no-significant-hazards 
consideration.
    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on that review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the request for amendments involves NSHC.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: January 22, 2008.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) 3.8.3 requirements related to 
Diesel Fuel Oil, Lube Oil, and Starting Air by replacing the specific 
fuel oil and lube oil storage values with the corresponding number of 
days supply. The specific volumes would be relocated to a licensee-
controlled document (i.e., the TS Bases). It would also expand the 
``clear and bright'' test in TS 5.5.10 by allowing a water and sediment 
test to be performed to establish the acceptability of new fuel oil 
prior to addition to the storage tanks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the Diesel Fuel Oil, Lube Oil, and 
Starting Air Specification relocates the volume of diesel fuel oil 
and lube oil required to support 7 day operation of the onsite 
diesel generators, and the volume equivalent to a 6 day supply, to 
licensee control. The specific volume of fuel oil equivalent to a 7 
and 6 day supply is calculated using the NRC approved methodology 
described in Regulatory Guide 1.137, Revision 1, ``Fuel Oil Systems 
for Standby Diesel Generators'' and ANSI/ANS [American National 
Standards Institute/American Nuclear Society] 59.51-1997 (formerly 
ANSI N195-1976), ``Fuel Oil Systems for Safety-Related Emergency 
Diesel Generators.'' The specific volume of lube oil equivalent to a 
7 and 6 day supply is based on the Emergency Diesel Generator (EDG) 
manufacturer's consumption values for the run time of the EDG. 
Because the requirements to maintain a 7 day supply of diesel fuel 
oil and lube oil are not changed and are consistent with the 
assumptions in the accident analyses, and the actions taken when the 
volume of fuel oil and lube oil are less than a 6 day supply have 
not changed, neither the probability nor the consequences of any 
accident previously evaluated will be affected. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the Diesel Fuel Oil Testing Program adds 
an option to use already approved testing methodology. Since the 
methodology is already discussed in ASTM D975 [``Standard 
Specification for Diesel Fuel Oils''] as an acceptable standard to 
determine water and sediment content, neither the probability nor 
the consequences of any accident previously evaluated will be 
affected. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the Diesel Fuel Oil, Lube Oil and 
Starting Air Specification and Diesel Fuel Oil Testing Program do 
not involve physical alterations of the plant (i.e., no new or 
different type of equipment will be installed) or changes in the 
methods governing normal plant operation. The changes do not alter 
assumptions made in the safety analysis but ensure that the diesel 
generator operates as assumed in the accident analysis. The proposed 
changes are consistent with the safety analysis assumptions. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the Diesel Fuel Oil, Lube Oil, and 
Starting Air Specification relocates the volume of diesel fuel oil 
and lube oil required to support 7 day operation of the onsite 
diesel generators, and the volume equivalent to a 6 day supply, to

[[Page 25038]]

licensee control. As the bases for the existing limits on diesel 
fuel oil and lube oil are not changed and the methods used to 
determine these limits have been previously approved, no change is 
made to the accident analysis assumptions and no margin of safety is 
reduced as part of this change. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.
    The proposed change to the Diesel Fuel Oil Testing Program 
provides an option to use a quantitative method of testing for 
sediment and water content as an alternative to a qualitative 
method. This option uses an already accepted method for assessing 
fuel oil quality. Based on this, there are no alterations to any 
assumptions used in the accident analysis and this change does not 
reduce any margin of safety. 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: 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. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

    Date of amendment request: February 7, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) Surveillance Requirement (SR) 
3.1.3.2 frequency in TS 3.1.3, ``Control Rod OPERABILITY'' from ``7 
days after the control rod is withdrawn and THERMAL POWER is greater 
than the [Low Power Setpoint] LPSP of [Rod Worth Minimizer] RWM'' to 
``31 days after the control rod is withdrawn and THERMAL POWER is 
greater than the LPSP of the RWM'' and revise Example 1.4-3 in Section 
1.4 ``Frequency'' to clarify the applicability of the 1.25 surveillance 
test interval extension. The proposed amendment does not adopt the 
clarification of Source Range Monitor (SRM) TS action for inserting 
control rods. This clarification was previously adopted during the 
JAFNPP conversion to Improved Standard Technical Specifications, TS 
Section 3.3.1.2, required Action E.2, ``Source Range Monitoring [SRM] 
Instrumentation.''
    Date of publication of individual notice in Federal Register: April 
2, 2008 (73 FR 18008).
    Expiration date of individual notice: May 2, 2008.

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

    Date of amendment request: March 13, 2008.
    Description of amendment request: The licensee proposes to change 
the Surveillance Requirement (SR) 3.6.5.8 to require verification that 
the reactor building spray nozzles are unobstructed following 
maintenance that could result in nozzle blockage in lieu of the current 
SR of performing the test every 10 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Reactor Building Spray System is not an initiator of any 
analyzed event. The proposed change does not have a detrimental 
impact on the integrity of any plan structure, system, or component 
that may initiate an analyzed event. The proposed change will not 
alter the operation or otherwise increase the failure probability of 
any plant equipment that can initiate an analyzed accident. This 
change does not affect the plant design. There is no increase in the 
likelihood of formation of significant corrosion products. Due to 
their location at the top of the containment, introduction of 
foreign material into the spray headers is unlikely. Foreign 
materials exclusion controls during and following maintenance 
provides assurance that the nozzles remain unobstructed. 
Consequently, there is no significant increase in the probability of 
an accident previously evaluated.
    The Reactor Building Spray system is designed to address the 
consequences of a Loss of Coolant Accident (LOCA) or a Main 
Steamline Break (MSLB) inside the reactor building. The Reactor 
Building Spray system is capable of performing its function 
effectively with the single failure of any active component in the 
system, any of its subsystems, or any of its support systems.
    Therefore, the consequences of an accident previously evaluated 
are not significantly affected by the proposed change.
    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 will not physically alter the plant (no new 
or different type of equipment will be installed) or change the 
methods governing normal 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 system piping and nozzles are made if material that is not 
susceptible to corrosion. Obstruction from sources external to the 
system is highly unlikely due to the location high in the reactor 
building and not being readily accessible. Strict controls are 
established to ensure the foreign material is not introduced into 
the Reactor Building Spray system during maintenance or repairs. 
Maintenance activities that could introduce significant foreign 
material into the system require subsequent system cleanliness 
verification which would prevent nozzle blockage. The spray header 
nozzles are expected to remain unblocked and available in the event 
that the safety function is required. The capacity of the system 
would remain unaffected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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--Entergy Nuclear Operations, P.O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: March 13, 2008.
    Description of amendment request: The proposed changes would 
replace the current Technical Specification (TS) 3.4.12, ``RCS [Reactor 
Coolant System] Specific Activity'' limit on reactor coolant system 
(RCS) gross specific activity with a new limit on RCS noble gas 
specific activity. The noble gas specific activity limit would be based 
on a new dose equivalent Xe-133 (DEX) definition that would replace the 
current E Bar average disintegration energy definition. In addition, 
the current dose equivalent I-131 (DEI) definition would be revised to 
allow the use of additional thyroid dose conversion factors (DCFs).
    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

[[Page 25039]]

consequences of an accident previously evaluated?
    Response: No.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the current Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously calculated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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--Entergy Nuclear Operations, P.O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: March 13, 2008.
    Description of amendment request: The proposed changes would 
replace the current TS 3.4.8, ``Reactor Coolant System Specific 
Activity'' limit on reactor coolant system (RCS) gross specific 
activity with a new limit on RCS noble gas specific activity. The noble 
gas specific activity limit would be based on a new dose equivalent Xe-
133 (DEX) definition that would replace the current E Bar average 
disintegration energy definition. In addition, the current dose 
equivalent I-131 (DEI) definition would be revised to allow the use of 
additional thyroid dose conversion factors (DCFs).
    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.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the current Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously calculated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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--Entergy Nuclear Operations, P. O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: March 13, 2008.
    Description of amendment request: The proposed change will relocate 
Technical Specification (TS) 3.4.7, ``Reactor Coolant System 
Chemistry,'' to the Technical Requirements Manual (TRM).
    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 acts to relocate current Reactor Coolant 
System (RCS) chemistry limits and monitoring requirements from the 
TSs to the TRM. Monitoring and maintaining RCS chemistry minimizes 
the potential for corrosion of RCS piping and components. Corrosion 
effects are considered a long-term impact on RCS structural 
integrity. Because RCS chemistry will continue to be monitored and 
controlled, relocating the current TS requirements to the TRM will 
not present an adverse impact to the RCS and, subsequently, will not 
impact the probability or consequences of an accident previously 
evaluated. Furthermore, once relocated to the TRM, changes to RCS 
chemistry limits or monitoring requirements will be controlled in 
accordance with 10 CFR 50.59.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 25040]]

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 does not result in any plant modifications 
or changes in the way the plant is operated. The proposed change 
only acts to relocate current RCS chemistry limits and monitoring 
requirements from the TSs to the TRM.
    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 will maintain limits on RCS chemistry 
parameters and will continue to provide associated monitoring 
requirements. Once relocated to the TRM, changes to RCS chemistry 
limits or monitoring requirements will be controlled in accordance 
with 10 CFR 50.59. In addition, the RCS chemistry limits are not a 
structure, system, or component which operating experience or 
probabilistic risk assessment has shown to be significant to public 
health and safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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--Entergy Nuclear Operations, P.O. Box 31995, Jackson, 
Mississippi 39286-1995.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: December 12, 2007.
    Description of amendment request: The proposed changes are 
administrative in nature and provide editorial changes to the technical 
specifications (TSs). The proposed changes involve: (1) Correcting the 
index; (2) removing cycle specific requirements or notes that have 
since expired and are no longer applicable; (3) deleting references to 
previously deleted requirements; (4) changing references to the 
location of previously relocated information; and (5) other editorial 
corrections. These proposed changes correct minor inconsistencies that 
have been introduced over time as a result of previous changes to the 
TSs or involve changes that are solely editorial in nature.
    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 proposed changes are administrative in nature and 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 mitigation of accidents or transient 
events.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
alter plant configuration, require that new plant equipment be 
installed, alter assumptions made about accidents previously 
evaluated, or impact the function of plant SSCs or the manner in 
which SSCs are operated, maintained, modified, tested, or inspected. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are administrative in nature and 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 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS),

    Units 2 and 3, York and Lancaster Counties, Pennsylvania 
    Date of amendment request: July 13, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to support application of 
Alternative Source Term (AST) methodology at PBAPS Units 2 and 3. The 
fission product release from the reactor core into containment is 
referred to as the ``source term,'' and is characterized by the 
composition and magnitude of the radioactive material, the chemical and 
physical properties of the material, and the timing of the release from 
the reactor core as discussed in Technical Information Document (TID) 
14844, ``Calculation of Distance Factors for Power and Test Reactor 
Sites.'' Since the publication of TID 14844, advances have been made in 
understanding the composition and magnitude, chemical form, and timing 
of fission product releases from severe nuclear power plant accidents. 
In light of these insights, NUREG-1465, ``Accident Source Terms for 
Light-Water Nuclear Power Plants,'' was published in 1995 with revised 
ASTs for use in the licensing of future light-water reactors.
    The Nuclear Regulatory Commission (NRC), in Title 10 of the Code of 
Federal Regulations, Section 50.67 (10 CFR 50.67), ``Accident source 
term,'' subsequently allowed the use of the ASTs described in NUREG-
1465 at operating plants. This request to apply the AST methodology is 
made in accordance with 10 CFR 50.67, with the exception that TID 14844 
will continue to be used as the radiation dose basis for equipment 
qualification at PBAPS Units 2 and 3. Application of the AST 
methodology at PBAPS Units 2 and 3 requires that radiation dose limits 
specified in 10 CFR 50.67 are adhered to for the exclusion area 
boundary, the low population zone outer boundary, and the facility 
control room.
    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

[[Page 25041]]

consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The implementation of alternative source term (AST) assumptions 
has been evaluated in revisions to the analyses of the following 
limiting design basis accidents (DBAs) at Peach Bottom Atomic Power 
Station (PBAPS):
     Loss-of-Coolant Accident,
     Fuel Handling Accident,
     Control Rod Drop Accident, and
     Main Steam Line Break Accident.
    Based upon the results of these analyses, it has been 
demonstrated that, with the requested changes, the dose consequences 
of these limiting events are within the regulatory guidance provided 
by the NRC for use with the AST. This guidance is presented in 10 
CFR 50.67 and associated Regulatory Guide 1.183, and Standard Review 
Plan Section 15.0.1. The Alternative Source Term is an input to 
calculations used to evaluate the consequences of an accident, and 
does not by itself affect the plant response, or the actual pathway 
of the radiation released from the fuel. It does, however, better 
represent the physical characteristics of the release, so that 
appropriate mitigation techniques may be applied. Therefore, the 
consequences of an accident previously evaluated are not 
significantly increased.
    The equipment affected by the proposed changes is mitigative in 
nature, and relied upon after an accident has been initiated. 
Application of the Alternative Source Term (AST) does not involve 
any physical changes to the plant design. While the operation of 
various systems do change as a result of these proposed changes, 
these systems are not accident initiators. Application of the AST is 
not an initiator of a design basis accident. The proposed changes to 
the Technical Specifications (TS), while they revise certain 
performance requirements, do not involve any physical modifications 
to the plant. As a result, the proposed changes do not affect any of 
the parameters or conditions that could contribute to the initiation 
of any accidents. As such, removal of operability requirements 
during the specified conditions will not significantly increase the 
probability of occurrence for an accident previously analyzed. Since 
design basis accident initiators are not being altered by adoption 
of the Alternative Source Term analyses, the probability of an 
accident previously evaluated is not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed changes). Similarly, it does not 
physically change any structures, systems or components involved in 
the mitigation of any accidents; thus, no new initiators or 
precursors of a new or different kind of accident are created. New 
equipment or personnel failure modes that might initiate a new type 
of accident are not created as a result of the proposed amendment.
    As such, the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Safety margins and analytical conservatisms have been evaluated 
and have been found acceptable. The analyzed events have been 
carefully selected and margin has been retained to ensure that the 
analyses adequately bound postulated event scenarios. The dose 
consequences due to design basis accidents comply with the 
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide 
1.183. The proposed amendment is associated with the implementation 
of a new licensing basis for PBAPS Design Basis Accidents (DBAs). 
Approval of the change from the original source term to a new source 
term taken from Regulatory Guide 1.183 is being requested. The 
results of the accident analyses, revised in support of the proposed 
license amendment, are subject to revised acceptance criteria. The 
analyses have been performed using conservative methodologies, as 
specified in Regulatory Guide 1.183. Safety margins have been 
evaluated and analytical conservatism has been utilized to ensure 
that the analyses adequately bound the postulated limiting event 
scenario. The dose consequences of these DBAs remain within the 
acceptance criteria presented in 10 CFR 50.67, ``Accident Source 
Term'', and Regulatory Guide 1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area boundary (EAB) and low population zone boundary 
(LPZ), as well as the Control Room, are within corresponding 
regulatory limits.
    Therefore, operation of PBAPS in accordance with the proposed 
changes will 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. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: March 31, 2008.
    Description of amendment request: FPL Energy Point Beach, LLC, 
requests adoption of an approved change to the Standard Technical 
Specifications (STS) for pressurized-water reactor (PWR) plants (NUREG-
1430, NUREG-1431, & NUREG-1432) and plant-specific technical 
specifications (TS), to replace the current limits on primary coolant 
gross specific activity with limits on primary coolant noble gas 
activity. The noble gas activity would be based on dose equivalent 
Xenon-133 and would take into account only the noble gas activity in 
the primary coolant. In addition, the current dose equivalent I-131 
definition would be revised to allow the use of additional thyroid dose 
conversion factors. The changes are consistent with Nuclear Regulatory 
Commission (NRC)-approved Industry/Technical Specification Task Force 
(TSTF) Standard Technical Specification Change Traveler, TSTF-490, 
Revision 0.
    Basis for proposed no-significant-hazards-consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary coolant 
gross activity is not within limit is not an initiator for any accident 
previously evaluated. The current variable limit on primary coolant 
iodine concentration is not an initiator to any accident previously 
evaluated. As a result, the proposed change does not significantly 
increase the probability of an accident. The proposed change will limit 
primary coolant noble gases to concentrations consistent with the 
accident analyses. The proposed change to the Completion Time has no 
impact on the consequences of any design basis accident since the 
consequences of an accident during the extended Completion Time are the 
same as the consequences of an accident during the Completion Time. As 
a result, the consequences of any accident previously evaluated are not 
significantly increased.
    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 in specific activity limits does not alter any 
physical part of the plant nor does it affect any plant operating 
parameter.

[[Page 25042]]

The change does not create the potential for a new or different kind of 
accident from any previously calculated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change revises the limits on noble gas radioactivity 
in the primary coolant. The proposed change is consistent with the 
assumptions in the safety analyses and will ensure the monitored values 
protect the initial assumptions in the safety analyses. Based upon the 
reasoning presented above, the requested change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the 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: Antonio Fernandez, Esquire, Senior Attorney, 
FPL Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-
0420.
    NRC Branch Chief: Lois M. James.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: March 31, 2008.
    Description of amendment request: The licensee proposed to increase 
the current maximum power level authorized by Section 2.C(1) of the 
renewed facility operating license from 1,775 megawatts thermal (Mwt) 
to 1,870 Mwt, an approximately five percent increase from the current 
licensed thermal power. The current maximum power level of 1,775 Mwt 
was approved in 1998, an increase of 6.3 percent from the original 
licensed thermal power of 1670 Mwt. Thus, when approved, the licensee's 
proposed amendment would take the maximum power level to about 12 
percent above the original license thermal power. The licensee's 
application addresses in details each of the following major technical 
areas: Extended power uprate, containment analysis methods change, 
increase in credit for containment overpressure for low head emergency 
core cooling system (ECCS) pumps, and reactor internal pressure 
differentials (RIPDs) for the steam dryer.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC). 
The licensee's NSHC analysis, addressing each technical area listed 
above, is reproduced below:

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

Extended Power Uprate

    Response: No.
    The probability (frequency of occurrence) of [d]esign [b]asis 
[a]ccidents occurring is not affected by the increased power level, 
because Monticello Nuclear Generating Plant (MNGP) continues to 
comply with the regulatory and design basis criteria established for 
plant equipment. A probabilistic risk assessment demonstrates that 
the calculated core damage frequencies do not significantly change 
due to [e]xtended [p]ower [u]prate (EPU). Scram setpoints (equipment 
settings that initiate automatic plant shutdowns) are established 
such that there is no significant increase in scram frequency due to 
EPU. No new challenges to safety-related equipment result from EPU.
    The changes in consequences of postulated accidents, which would 
occur from 102 percent of the EPU [rated thermal power] RTP compared 
to those previously evaluated, are acceptable. The results of EPU 
accident evaluations do not exceed the NRC[-] approved acceptance 
limits. The spectrum of postulated accidents and transients has been 
investigated, and are shown to meet the plant's currently licensed 
regulatory criteria. In the area of fuel and core design, for 
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and 
other applicable Specified Acceptable Fuel Design Limits (SAFDL) are 
still met. Continued compliance with the SLMCPR and other SAFDLs 
will be confirmed on a cycle[-]specific basis consistent with the 
criteria accepted by the NRC.
    Challenges to the [r]eactor [c]oolant [p]ressure [b]oundary were 
evaluated at EPU conditions (pressure, temperature, flow, and 
radiation) and were found to meet their acceptance criteria for 
allowable stresses and overpressure margin. Challenges to the 
containment have been evaluated, and the containment and its 
associated cooling systems continue to meet the current licensing 
basis. The increase in the calculated post[-] LOCA suppression pool 
temperature above the currently assumed peak temperature was 
evaluated and determined to be acceptable. Radiological release 
events (accidents) have been evaluated, and have been shown to meet 
the guidelines of 10 CFR 50.67.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR [residual heat 
removal] heat exchanger capability K-value, and mechanistic heat and 
mass transfer from the suppression pool surface to the wetwell 
airspace after 30 seconds for the long[-]term design[-] basis [-
]accident loss of coolant accident (DBA-LOCA) containment analysis 
are not relevant to accident initiation, but rather, pertain to the 
method used to accurately evaluate postulated accidents. The use of 
these elements does not, in any way, alter existing fission product 
boundaries, and provides a conservative prediction of the 
containment response to DBA-LOCAs. Therefore, the containment 
analysis method change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Increase in Credit for Containment Overpressure for Low Head 
Emergency Core Cooling System (ECCS) Pumps

    Response: No.
    These changes update parameters used in the MNGP safety analyses 
and expand the range and scope of the analyses. This will result in 
a more realistic analysis of available containment overpressure 
under design [-]basis accident conditions. The updated analyses 
affect only the evaluation of previously reviewed accidents. No 
plant structure, system, or component (SSC) is physically affected 
by the updated and expanded analyses. No method of operation of any 
plant SSC is affected. Therefore, there is no significant increase 
in the probability or consequence of a previously evaluated 
accident.

Reactor Internal Pressure Differentials (RIPDs) for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The values more accurately 
represent the actual plant configuration. No plant structure, 
system, or component (SSC) is physically affected by the updated and 
expanded analyses. No method of operation of any plant SSC is 
affected. Therefore, there is no significant increase in the 
probability or consequence of a previously evaluated accident.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt, which bounds this license 
amendment request to operate at 1,870 Mwt. 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?

Extended Power Uprate

    Response: No.
    Equipment that could be affected by EPU has been evaluated. No 
new operating mode, safety-related equipment lineup, accident 
scenario, or equipment failure mode was identified. The full 
spectrum of accident considerations has been evaluated and no new or 
different kind of accident has been identified. EPU uses developed 
technology and applies it within capabilities of existing or 
modified plant safety[-]related equipment in accordance with the 
regulatory criteria (including NRC[-]approved codes, standards and 
methods). No new accidents or event precursors have been identified.
    The MNGP TS require revision to implement EPU. The revisions 
have been assessed and it was determined that the proposed change 
will not introduce a different accident than that previously 
evaluated. Therefore, the proposed changes

[[Page 25043]]

do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR heat exchanger 
capability K-value, and mechanistic heat and transfer from the 
suppression pool surface to the wetwell airspace after 30 seconds 
for the long term DBA-LOCA containment analysis are not relevant to 
accident initiation, but pertain to the method used to evaluate 
currently postulated accidents. The use of these analytical tools 
does not involve any physical changes to plant structures or 
systems, and does not create a new initiating event for the spectrum 
of events currently postulated. Further, they do not result in the 
need to postulate any new accident scenarios. Therefore, the 
containment analysis method change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

Increase in Credit for Containment Overpressure for Low Head ECCS 
Pumps

    Response: No.
    The proposed change involves the updating and expansion in scope 
of the existing design bases analysis with respect to the available 
containment overpressure. No new failure mode or mechanisms have 
been created for any plant SSC important to safety nor has any new 
limiting single failure been identified as a result of the proposed 
analytical changes. Therefore, the change to containment 
overpressure credited for low pressure ECCS pumps does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.

Reactor Internal Pressure Differentials for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The steam dryer RIPDs are not 
relevant to accident initiation, but only pertain to the method used 
to evaluate reactor vessel internals loads. The revised steam dryer 
RIPD values more accurately represent the actual plant 
configuration. Therefore, the change to steam dryer RIPDs does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt, which bounds this license 
amendment request to operate at 1,870 Mwt. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?

Extended Power Uprate

    Response: No.
    The EPU affects only design and operational margins. Challenges 
to the fuel, reactor coolant pressure boundary, and containment were 
evaluated for EPU conditions. Fuel integrity is maintained by 
meeting existing design and regulatory limits. The calculated loads 
on affected structures, systems and components, including the 
reactor coolant pressure boundary, will remain within their design 
allowables for design[-]basis event categories. No NRC acceptance 
criterion is exceeded. Because the MNGP configuration and responses 
to transients and postulated accidents do not result in exceeding 
the presently approved NRC acceptance limits, the proposed changes 
do not involve a significant reduction in a margin of safety.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, variable RHR heat exchanger 
capability K-value, and mechanistic heat and mass transfer from the 
suppression pool surface to the wetwell airspace after 30 seconds 
for the long[-]term DBA-LOCA containment analysis are realistic 
phenomena and provide a conservative prediction of the plant 
response to DBA-LOCAs. The increase in pressure and temperature are 
relatively small and are within design limits. Therefore, the 
containment analysis methods change does not involve a significant 
reduction in the margin of safety.

Increase in Credit for Containment Overpressure for Low Head ECCS 
Pumps

    Response: No.
    The proposed changes revise containment response analytical 
methods and scope for containment pressure to assist in ECCS pump 
net positive suction head (NPSH). The changes are still based on 
conservative but more realistic analysis of available containment 
overpressure determined using analysis methods that minimize 
containment pressure and maximize suppression pool temperature. 
These changes do not constitute a significant reduction in the 
margin of safety.

Reactor Internal Pressure Differentials for the Steam Dryer

    Response: No.
    The revised steam dryer RIPDs are used in evaluating loads in 
reactor vessel internals for various conditions (i.e., during 
normal, upset and faulted conditions). The revised steam dryer RIPD 
values more accurately represent the actual plant configuration. The 
changes are still conservative but more accurately represent the 
MNGP configuration. These changes do not constitute a significant 
reduction in the margin of safety.
    The analyses supporting the above evaluations were performed at 
the EPU power level of 2,004 Mwt, which bounds this license 
amendment request to operate at 1,870 Mwt. 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 
the NRC staff's own analysis above, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: April 3, 2008.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements related to control 
room envelope (CRE) habitability in TS Section 3.7.4, ``Control Room 
Emergency Filtration (CREF) System,'' and Section 5.5, ``Programs and 
Manuals.'' The proposed changes are consistent with Technical 
Specification Task Force (TSTF) Standard Technical Specifications (STS) 
change TSTF-448, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC) by 
referencing the NRC staff's model NSHC analysis published on January 
17, 2007 (72 FR 2022). The NRC staff's model NSHC analysis is 
reproduced below:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of

[[Page 25044]]

design basis accidents. Thus, the consequences of any accident 
previously evaluated are not increased. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

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

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

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's referenced analysis, and 
has found that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the proposed 
amendment involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.

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: March 28, 2008.
    Description of amendment request: The amendments would revise PPL 
Susquehanna, LLC, Units 1 and 2 (PPL) Technical Specifications (TSs) 
3.8.4, ``DC Sources--Operating,'' to establish two new Conditions, A 
and B the associated Required Actions with their completion times, and 
also, make some editorial and administrative changes.
    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?
    No. The proposed changes revise the Technical Specifications 
(TS) for the DC Electrical Power Systems and propose new Actions 
with increased completion times for an inoperable battery charger. 
The DC electrical power systems, including associated battery 
chargers, are not initiators to any accident sequence analyzed in 
the Final Safety Analysis Report (FSAR). Operation in accordance 
with the proposed TS ensures that the DC electrical power systems 
are capable of performing functions as described in the FSAR. 
Therefore, the mitigative functions supported by the DC Power 
Systems will continue to provide the protection assumed by the 
analysis. The integrity of fission product barriers, plant 
configuration, and operating procedures as described in the FSAR 
will not be affected by the proposed changes.
    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?
    No. The proposed changes only involve revising the TS for the DC 
electrical power systems. The DC electrical power systems are used 
to supply equipment used to mitigate an accident. These mitigative 
functions, supported by the DC electrical power systems are not 
affected by these changes and they will continue to provide the 
protection assumed by the safety analysis described in the FSAR. 
There are no new types of failures or new or different kinds of 
accidents or transients that could be created by these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The margin of safety is established through equipment 
design, operating parameters, and the setpoints at which automatic 
actions are initiated. The proposed changes will not adversely 
affect operation of plant equipment. These changes will not result 
in a change to the setpoints at which protective actions are 
initiated. Sufficient DC electrical system capacity is ensured to 
support operation of mitigation equipment. The equipment fed by the 
DC electrical sources will continue to provide adequate power to 
safety related loads in accordance with the safety analysis.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    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: March 28, 2008.
    Description of amendment request: The amendments would revise PPL 
Susquehanna, LLC, Units 1 and 2 (PPL) Technical Specifications (TSs) TS 
3.6.4.1 ``Secondary Containment,'' and TS 3.6.4.3 ``Standby Gas 
Treatment System,'' as follows:
    (1) To add a new Required Action option for TS 3.6.4.1 Condition A, 
to allow additional time to restore secondary containment to OPERABLE 
when the inoperability is not caused by a loss of secondary containment 
integrity,
    (2) To add a new Actions note TS 3.6.4.1, to allow opening of 
secondary containment heating ventilation and air conditioning duct 
access doors and opening of a secondary containment equipment ingress/
egress door (102 door) under administrative controls provided no 
movement of irradiated fuel assemblies in the secondary containment, 
CORE ALTERATIONS, or operations with a potential for draining the 
reactor vessel (OPDRVs) are in progress,
    (3) To modify the existing note to Surveillance Requirement (SR) 
3.6.4.1.3 and add a second note to this same SR, to expand upon the 
existing SR exception note by adding other types of door access 
openings that occur for entry and exit of people or equipment, and
    (4) The administrative change to remove a one-time allowance in TS 
3.6.4.1 and TS 3.6.4.3 ``Standby Gas Treatment System [SGTS],'' that 
extended the allowable Completion Time for Secondary Containment

[[Page 25045]]

inoperable and two SGTS subsystems inoperable in MODE 1, 2, or 3. This 
allowance was previously incorporated into both Unit 1 and Unit 2 TSs 
to facilitate Reactor Recirculating Fan Damper Motor work.
    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.
    These changes do not involve any physical change to structures, 
systems, or components (SSCs) and do not alter the method of 
operation of any SSCs. The current assumptions in the safety 
analysis regarding accident initiators and mitigation of accidents 
are unaffected by these changes. No SSC failure modes or mechanisms 
are being introduced, and the likelihood of previously analyzed 
failures remains unchanged.
    Operation in accordance with the proposed new Required Action 
option for TS 3.6.4.1 Condition A and the Notes that are being 
modified and added in both the Unit 1 and Unit 2 Technical 
Specifications ensures that the secondary containment remains 
capable of performing its function. The Required Action change, 
which will permit up to 72 hours to restore secondary containment 
vacuum, only provides this additional time when it can be shown that 
the vacuum loss has not been caused through compromise of the 
secondary containment boundary.
    The proposed Note modifications and additions addressing 
secondary containment access door and duct access door openings will 
provide relief from TS requirements that must currently be 
implemented in response to various routine plant activities. These 
activities can be managed through administrative controls that will 
ensure doors can be closed quickly (within 30 minutes) to re-
establish secondary containment before the early in-vessel release 
phase begins (Regulatory Guide 1.183).
    These changes do not, therefore, result in an increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of any 
plant equipment. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints, at which protective or mitigative actions are 
initiated, affected by this change. This change does not alter the 
manner in which equipment operation is initiated, nor will the 
function demands on credited equipment be changed. No alterations in 
the procedures that ensure the plant remains within analyzed limits 
are being proposed, and no changes are being made to the procedures 
relied upon to respond to an off-normal event as described in the 
FSAR [final safety analysis report]. As such, no new failure modes 
are being introduced. The change does not alter assumptions made in 
the safety analysis and licensing basis.
    Therefore, 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.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes are acceptable because the 
Completion Time for the new Required Action to verify secondary 
containment boundary integrity within 4 hours has been established 
to be consistent with the current completion time of Condition A. A 
failure or inability to complete this verification will result in 
the implementation of LCO [limiting condition for operation] 3.6.4.1 
requirements in the same timeframe that currently exists. Upon 
successful completion of this verification, however, the proposed 
change will provide 72 hours to restore secondary containment to an 
operable status through vacuum restoration. When in this condition, 
the secondary containment and SGTS are capable of performing their 
design basis function.
    The Note modifications and additions to TS 3.6.4.1 are also 
acceptable because the revised Notes provide allowances and 
exemptions to Technical Specification entry for routine plant 
activities that can be administratively controlled and quickly 
restored.
    The plant response to analyzed events is not affected by these 
changes and will, continue to provide the margin of safety assumed 
by the safety analysis.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Mark G. Kowal.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment request: November 30, 2007.
    Description of amendment request: The proposed Technical 
Specification changes will provide operational flexibility supported by 
direct current (DC) electrical subsystem design upgrades that are in 
progress. These upgrades will provide increased capacity batteries, 
additional battery chargers, and the means to cross-connect DC 
subsystems while meeting all design battery loading requirements. With 
these modifications in place, it will be feasible to perform routine 
surveillances as well as battery replacements online.
    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?
    Response: No.
    The proposed changes to Technical Specifications (TS) 3.8.4 and 
3.8.6 would allow extension of the Completion Time (CT) for 
inoperable Direct Current (DC) distribution subsystems to manually 
cross-connect DC distribution buses of the same safety train of the 
operating unit for 21 days (30 days for upgrade to 1800 amp-hour 
rated batteries). Currently the CT only allows for 2 hours to 
ascertain the source of the problem before a controlled shutdown is 
initiated. Loss of a DC subsystem is not an initiator of an event. 
However, complete loss of a Train A (subsystems A and C) or Train B 
(subsystems B and D) DC system would initiate a plant transient/
plant trip.
    Operation of a DC Train in cross-connected configuration does 
not affect the quality of DC control and motive power to any system. 
Therefore, allowing the cross-connect of DC distribution systems 
does not significantly increase the probability of an accident 
previously evaluated in Chapter 15 of the Updated Final Safety 
Analysis Report (UFSAR).
    The above conclusion is supported by Probabilistic Risk 
Assessment (PRA) evaluation which encompasses all accidents, 
including UFSAR Chapter 15.
    New TS Surveillance Requirement (SR) 3.8.4.4 is added to allow 
the application of the modified performance discharge testing on 
batteries rated at 1800 amp-hour using a frequency of 30 months. The 
application of the modified performance test is the preferred choice 
at SONGS for Class 1 E 1800 amp-hour rated batteries. Therefore, 
only the modified performance discharge test will be used which uses 
the combined duty cycle of the cross-connected subsystems AC or B-D. 
Battery life expectancy is optimized by using a 30-month modified 
performance test (service and performance test combined). The more 
rigorous modified performance discharge test will be applied in 
intervals of 30 months over the entire battery life. Using the same 
test method and test frequency throughout the battery life ensures 
that best

[[Page 25046]]

trending results are achieved. The test frequency of 30 months will 
better correspond with scheduling of the more rigorous 60-month 
interval battery performance of modified performance discharge 
tests. Based on operating experience, the interval of 30 months is 
not expected to affect SONGS' capability to detect battery health 
and capacity.
    The relocation of preventive maintenance surveillances and 
certain operating limits and actions to the Licensee Controlled 
Specifications and new Battery Monitoring and Maintenance Program 
will not challenge the ability of the DC electrical power system to 
perform its design function. Appropriate monitoring and maintenance 
consistent with industry standards will continue to be performed. In 
addition, the DC electrical power system is within the scope of 10 
CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure the control 
of maintenance activities associated with the DC electrical power 
system. Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have 
been incorporated into TSs 3.8.4, 3.8.5, and 3.8.6. These changes do 
not impact the probability or consequences of an accident previously 
evaluated.
    Further, changes are made of an editorial nature or provide 
clarification regarding electrical `Trains' and `Subsystems' by 
using a more conventional terminology. TSs affected by editorial 
changes include 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, 3.8.9, and 
3.8.10. The changes being proposed in the TS do not affect 
assumptions contained in other safety analyses or the physical 
design of the plant, nor do they affect other Technical 
Specifications that preserve safety analysis assumptions.
    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 analyzed.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the UFSAR. Rather, the DC electrical power 
system is used to supply equipment used to mitigate an accident.
    The proposed change modifies TSs and surveillances for batteries 
and chargers to meet the improvements of TSTF-360, Rev. 1 and IEEE 
450-2002 whose intent is to maintain the same equipment capability 
as previously assumed in Southern California Edison's (SCE's) 
commitment to IEEE 450-1980.
    The proposed change will allow the cross-tie of DC subsystems 
and allow extension of the CT for an inoperable subsystem to 21 days 
(30 days for upgrade to 1800 amp-hour rated batteries). Failure of 
the cross-tied DC buses and/or associated battery(ies) is bounded by 
existing evaluations for the failure of an entire electrical train.
    Swing battery chargers are added to increase the overall DC 
system reliability. Administrative and mechanical controls are in 
place to ensure the design and operation of the DC systems continue 
to meet the UFSAR design basis.
    Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of new or different 
kind of accident from any accident 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new battery maintenance 
and monitoring program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to safety 
related loads in accordance with analysis assumptions.
    Improvements in accordance with IEEE 450-2002 and TSTF-360, Rev. 
1 maintain the same level of equipment performance stated in the 
UFSAR and the current Technical Specifications.
    The addition of swing battery chargers increases the overall DC 
system reliability. Administrative and mechanical controls will be 
in place to ensure that the design and operation of the DC systems 
continue to meet the UFSAR design basis.
    The addition of the DC cross-tie capability proposed for TS 
3.8.4 has been evaluated, as described previously, using PRA and 
determined to be of acceptable risk as long as the duration while 
cross-tied is limited to 30 days. A new Condition has been included 
as part of this proposed change to ensure that plant operation, with 
DC buses cross-tied, will not exceed 21 days (30 days for upgrade to 
1800 amp-hour rated batteries).
    Revising the LCO statement to reflect the SONGS-specific design 
terminology and renaming existing conditions to make the Condition 
more consistent with the Standard Technical Specifications (STS) is 
considered administrative.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: August 29, 2006, as supplemented 
November 6, November 27, 2006, January 30, June 22, July 16, August 13, 
October 18, December 11, 2007, January 24, February 4, February 25 (two 
letters, nos. 1389 and 0175), February 27, and March 13, 2008.
    Description of amendment request: The proposed amendments would 
revise the licensing and design basis, including the Technical 
Specifications, with a full scope implementation of an alternative 
source term (AST). The licensee states that the AST analyses include 
determination of the onsite radiological doses, specifically the main 
control room, technical support center and off-site radiological doses 
resulting from the loss-of-coolant, main steam line break, control rod 
drop, and fuel-handling design-basis accident (DBA) analyses. The 
licensee states that the analyses demonstrate that, using AST 
methodologies, the post-accident onsite and offsite doses remain within 
regulatory acceptance limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Adoption of the AST and those plant systems affected by 
implementing AST do not initiate DBAs. The AST does not affect the 
design or manner in which the facility is operated; rather, once the 
occurrence of an accident has been postulated, the new accident 
source term is an input to analyses that evaluate the radiological 
consequences. The implementation of the AST and changed Technical 
Specifications have been incorporated in the analyses for the 
limiting DBAs at HNP. The structures, systems, and components 
affected by the proposed change are mitigative in nature and relied 
upon after an accident has been initiated. Based on the revised 
analyses, the proposed changes to the Technical Specifications 
(including revised leakage limits) impose certain performance 
criteria which do not increase accident initiation probability. The 
proposed changes do not involve a revision to the parameters

[[Page 25047]]

or conditions that could contribute to the initiation of a DBA 
discussed in Chapter 15 of the Unit 2 Final Safety Analysis Report. 
Therefore, the proposed change does not result in an increase in the 
probability of an accident previously identified. Plant specific AST 
radiological analyses have been performed and, based on the results 
of these analyses, it has been demonstrated that the dose 
consequences of the limiting events considered in the analyses are 
within the regulatory guidance provided by the Nuclear Regulatory 
Commission for use with the AST. This guidance is presented in 
[Title 10 of the Code of Federal Regulations, Section 50.67] (10 CFR 
50.67), [Accident Source Term] Regulatory Guide 1.183, [Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors (ML003716792)] and Standard Review Plan, 
Section 15.0.1. Therefore, the proposed change does not result in a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Implementation of AST and associated changes does not alter or 
involve any design basis accident initiators. These changes do not 
affect the design function or mode of operations of systems, 
structures, or components in the facility prior to a postulated 
accident. Since systems, structures, and components are operated 
essentially no differently after the AST implementation, no new 
failure modes are created by this proposed change. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant decrease in 
the margin of safety?
    The changes proposed are associated with a revision to the 
licensing basis for HNP. Approval of the licensing basis change from 
the original source term to the AST is requested by this application 
for a license amendment. The results of the accident analyses 
revised in support of the proposed change are subject to the 
acceptance criteria in 10 CFR 50.67. The analyzed events have been 
carefully selected, and the analyses supporting these changes have 
been performed using approved methodologies and conservative inputs 
to ensure that analyzed events are bounding and safety margin has 
been retained. The dose consequences of these limiting events are 
within the acceptance criteria presented in 10 CFR 50.67, Regulatory 
Guide 1.183, and Standard Review Plan 15.0.1. Therefore, because the 
proposed changes continue to result in dose consequences within the 
applicable regulatory limits, the changes are considered to not 
result in 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Branch Chief: Melanie C. Wong.

Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: March 27, 2008.
    Description of amendment request: The proposed amendment would 
revise the allowable value for Function 3, ``Containment Purge Exhaust 
Radiation Monitors,'' in Technical Specifications (TSs) Table 3.3.6-1, 
``Containment Vent Isolation Instrumentation,'' of Limiting Conditions 
for Operation 3.3.6, during Modes 1 through 4. The current allowable 
value was found to be non-conservative for operating Modes 1 through 4 
because the basis for the specified value inappropriately credited the 
containment purge exhaust filters, which are only required during 
movement of irradiated fuel assemblies within containment. The current 
allowable value remains acceptable during movement of irradiated fuel 
assemblies within containment.
    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 is associated with radiation effluent 
monitoring and isolation of Containment Purge exhaust flow in the 
event of a design basis SBLOCA [small break loss of coolant 
accident]. The change is not associated with equipment or processes 
which can initiate a design basis accident. Consequently, this 
change does not affect the probability of an accident previously 
evaluated.
    The revised purge exhaust monitor allowable value will ensure 
the monitors isolate the purge exhaust and will limit the offsite 
doses associated with a SBLOCA to well within the limits of 10 CFR 
100. This change serves to ensure the consequences of an accident 
previously evaluated remain bounded by the plant's current licensing 
basis. Therefore, the consequences of accidents previously evaluated 
are not increased by this change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is associated with radiation effluent 
monitoring and isolation of Containment Purge exhaust flow in the 
event of a design basis SBLOCA. The change is not associated with 
equipment or processes which can initiate a design basis accident. 
The change does not introduce new accident initiators or physical 
changes in plant equipment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change involves a conservative change in the 
Containment Purge exhaust radiation monitor allowable value in TS 
Table 3.3.6-1. The new allowable value reflects a change in the 
monitor analytical limit which does not assume credit for the 
Containment Purge exhaust filters. The proposed allowable value will 
ensure the monitors will isolate the purge exhaust as assumed in the 
existing design basis SBLOCA analysis.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these

[[Page 25048]]

amendments satisfy the criteria for categorical exclusion in accordance 
with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no 
environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of application for amendment: April 12, 2007.
    Brief description of amendment: The amendment modifies the TMI-1 
technical specifications related to control room envelope habitability 
consistent with Technical Specification Task Force (TSTF) Traveler 
TSTF-448.
    Date of issuance: April 16, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 264.
    Facility Operating License No. DPR-50. Amendment revised the 
license and the technical specifications.
    Date of initial notice in Federal Register: June 5, 2007 (72 FR 
31100). The supplements dated January 18, 2008, and March 14, 2008, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed and did not 
change the NRC staff's original proposed no significant hazards 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 16, 2008.
    No significant hazards consideration comments received: No.

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 application for amendments: January 22, 2007, as 
supplemented on June 21, July 18, July 31, and October 15, 2007, and 
January 24, February 14, March 5, and March 21, 2008.
    Brief description of amendments: Change the Technical 
Specifications (TSs) to support the transition to AREVA fuel and core 
design methodologies.
    Date of issuance: March 27, 2008.
    Effective date: Date of issuance, to be implemented on Unit 1 prior 
to startup from the 2008 refueling outage, and to be implemented on 
Unit 2 prior to startup from the 2009 refueling outage.
    Amendment Nos.: 246 and 274.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the TSs.
    Date of initial notice in Federal Register: December 4, 2007 (72 FR 
68208). The supplements dated January 24, February 14, March 5, and 
March 21, 2008, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 27, 2008.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: November 7, 2007.
    Brief description of amendment: The amendment deletes License 
Condition 2.F, which requires reporting of violations of certain other 
requirements contained in Section 2.C of the license.
    Date of issuance: April 15, 2008.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 206.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: December 4, 2007 (72 FR 
68211) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 15, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 22, 2007.
    Brief description of amendment: The amendment incorporates 
technical specification (TS) changes based on Nuclear Regulatory 
Commission (NRC)-approved TS Task Force (TSTF)-497-A, ``Changes to 
Reflect Revision of 10 CFR 50.55a,'' Revision 0, as modified by NRC-
approved TSTF-497, ``Limit Inservice Testing Program [Surveillance 
Requirements] SR 3.0.2 Application to Frequencies of Two years or 
Less.'' Specifically, the amendment revises Palisades Nuclear Plant TS 
Section 5.5.7, ``Inservice Testing Program,'' to update references to 
the American Society of Mechanical Engineers code and applicability of 
the provisions of SR 3.0.2.
    Date of issuance: April 15, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 232.
    Renewed Facility Operating License No. DPR-20: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 28, 2007 (72 FR 
49575). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 15, 2008.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: October 18, 2007.
    Brief description of amendment: The amendment revised the Technical 
Specifications to change requirements related to emergency diesel 
generator (EDG) fuel oil tank volume, EDG fuel oil testing and reactor 
building crane inspections.
    Date of Issuance: April 17, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 231.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.

[[Page 25049]]

    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71711).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 17, 2008.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: April 24, 2007, as supplemented by 
electronic mail dated February 12, 2008.
    Brief description of amendment: The change adds Optimized ZIRLO as 
an acceptable fuel rod cladding material in the Waterford Steam 
Electric Station, Unit 3, Technical Specification (TS) 5.3.1, ``Fuel 
Assemblies.'' TS 5.3.1 currently identifies, in part, Zircaloy or ZIRLO 
\PM\ fuel rod cladding as the allowable fuel rod cladding material.
    Date of issuance: April 16, 2008.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 215.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 22, 2007 (72 FR 
28720). The supplemental electronic mail dated February 12, 2008, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 16, 2008.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: August 2, 2007, as supplemented by 
letters dated January 17, March 10, and electronic mail dated March 24, 
2008. In addition, Entergy submitted for review and approval the 
revised emergency core cooling system (ECCS) performance analysis by 
letter dated August 9, 2007, as supplemented by letter dated January 
21, 2008; and a supplement to the ECCS performance analysis by letter 
dated October 4, 2007, as supplemented by letter dated March 4, 2008.
    Brief description of amendment: The changes to the technical 
specifications add new analytical methods and modify the containment 
average air temperature and safety injection tank level to support the 
implementation of Combustion Engineering 16 x 16 Next Generation Fuel 
(NGF) as defined in Westinghouse Topical Report WCAP-16500-P beginning 
in Cycle 16 commencing after the spring 2008 refueling outage.
    Date of issuance: April 15, 2008.
    Effective date: As of the date of issuance and shall be shall be 
implemented prior to startup following the spring 2008 refueling 
outage.
    Amendment No.: 214.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2007 (72 
FR 51858). The supplemental letters dated January 17, and March 10, 
2008, and electronic mail dated March 24, 2008, for changes to the TSs; 
the supplemental letter dated January 21, 2008, for review and approval 
of the revised ECCS performance analysis; and the supplemental letter 
dated March 4, 2008, for review and approval of the supplement to the 
ECCS performance analysis, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 25, 2008, as 
supplemented by letters dated March 27, 2008, and April 9, 2008.
    Brief description of amendment: The amendments revise Technical 
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' and TS 
5.6.9, ``Steam Generator (SG) Tube Inspection Report.'' For TS 5.5.9, 
the amendment replaces the existing alternate repair criteria in the 
provisions for SG tube repair criteria, during Braidwood, Unit 2, 
Refueling Outage 13 and the subsequent operating cycle. For TS 5.6.9, 
three new reporting requirements are added to the existing seven 
requirements for Braidwood Station (Braidwood), Unit 2. These changes 
only affect Braidwood, Unit 2; however, this action is docketed for 
Braidwood, Units 1 and 2, because the TS are common to both units.
    Date of issuance: April 18, 2008.
    Effective date: As of the date of issuance and shall be implemented 
prior to the return to service from Braidwood, Unit 2, spring 2008 
Refueling Outage 13.
    Amendment Nos.: Unit 1-150; Unit 2-150.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendment 
revised the TSs and License.
    Date of initial notice in Federal Register: March 11, 2008 (73 FR 
13029).
    The March 27, 2008, and April 9, 2008, supplemental letters 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 18, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 29, 2008.
    Brief description of amendments: The amendments revised the 
licensing basis of ice condenser ice fusion time, specifying conditions 
under which plant operation may proceed in less than 5 weeks after ice 
baskets have been reloaded.
    Date of issuance: April 16, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented prior to Unit 1 entering Mode 4 at the end of the 2008 
refueling outage.
    Amendment No.: 303 (for DCCNP-1) and 286 (for DCCNP-2).
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Renewed Operating Licenses.
    Date of initial notice in Federal Register: March 12, 2008 (73 FR 
13253)
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated April 16, 2008.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendment: July 26, 2007, as supplemented 
by letters dated October 3 and December 21, 2007, and February 29, 
2008.

[[Page 25050]]

    Brief description of amendment: The proposed amendment would add a 
new reference to Technical Specification 6.9.1.14.a, which lists 
documents that have been approved by the U.S. Nuclear Regulatory 
Commission for use in determining the core operating limits. The new 
reference is the Areva NP, Inc., Topical Report EMF-2103P-A, 
``Realistic Large Break LOCA [Loss-Of-Coolant Accident] Methodology for 
Pressurized Water Reactors.''
    Date of issuance: April 10, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No. 311.
    Facility Operating License No. DPR-79: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: August 28, 2007 (72 FR 
49583). The supplemental letters dated October 3 and December 21, 2007, 
and February 29, 2008, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 2008.
    No significant hazards consideration comments received: No.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of amendment request: April 10, 2008.
    Brief Description of amendment request: The proposed amendment 
would revise Technical Specification (TS) 3.5.5, Refueling Water Tank 
(RWT) to increase the minimum required RWT level indications and the 
corresponding borated water volumes in TS Figure 3.5.5-1, ``Minimum 
Required RWT Volume,'' by 3 percent. This change will ensure that there 
is adequate water volume available in the RWT to ensure that the 
engineered safety feature pumps and the new containment recirculation 
sump strainers will meet their design functions during loss-of-coolant 
accidents.
    Date of publication of individual notice in Federal Register: April 
17, 2008 (73 FR 20961).
    Expiration date of individual notice: May 1, 2008.

    Dated at Rockville, Maryland, this 28th day of April, 2008.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
 [FR Doc. E8-9679 Filed 5-5-08; 8:45 am]
BILLING CODE 7590-01-P