Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 19106-19117 [E8-6904]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 68 (Tuesday, April 8, 2008)]
[Notices]
[Pages 19106-19117]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-6904]


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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 March 13, 2008 to March 26, 2008. The last 
biweekly notice was published on March 25, 2008 (73 FR 15780).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect

[[Page 19107]]

to issuance of the amendment to the subject facility operating license 
and any person whose interest may be affected by this proceeding and 
who wishes to participate as a party in the proceeding must file a 
written request via electronic submission through the NRC E-Filing 
system for a hearing and a petition for leave to intervene. Requests 
for a hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested person(s) should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition 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 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms Viewer\TM\ is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-
viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-
submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC

[[Page 19108]]

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 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: November 13, 2007.
    Description of amendment request: The proposed amendment would 
delete the Oyster Creek Technical Specification (TS) 6.5, ``Review and 
Audit.'' Specifically, the proposed change would delete TS 6.5.1, 
``Technical Review and Control,'' TS 6.5.2, ``Independent Safety Review 
Function,'' and TS 6.5.3, ``Audits'' which are currently being 
implemented by the Exelon/AmerGen Quality Assurance Topical Report. 
Additionally, the proposed amendment would correct typographical errors 
in Table 3.1.1, ``Protective Instrumentation Requirements'' and Table 
4.1.1, ``Minimum Check, Calibration and Test Frequency for Protective 
Instrumentation'' and would delete the Condenser Vacuum Pump Isolation 
Surveillance from Table 4.1.2, ``Minimum Test Frequencies for Trip 
Systems.'' The TS required operability associated with the Surveillance 
Requirement (SR) was removed from the Oyster Creek TSs via Amendment 
No. 169 and removal of the SR was inadvertently omitted. This request 
for approval of a license amendment was submitted concurrently with a 
similar request for Three Mile Island Nuclear Station, Unit No. 1 
(TMI). Due to some differences in the requested changes, the TMI 
amendment request will be noticed separately.
    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.
    No physical changes to the facilities, OCNGS [Oyster Creek] and 
TMI, will occur as a result of this proposed amendment. The proposed 
changes will not alter the physical design or operational procedures 
associated with any plant structure, system, or component.
    The proposed changes involve the deletion of several 
administrative requirements from the Technical Specifications (TS) 
that are now controlled under the Exelon/AmerGen Quality Assurance 
Topical Report (QATR) and several administrative procedures, AS-AA-
102, [Station Qualified Review Program], HU-AA-1212 (Independent 
Third Party Reviews), LS-AA-101 (License/TS changes), LS-AA-106 
(PORC), NO-AA-200-002 (Audits) and SY-AA-101-104 (Security Plan 
changes), and are, therefore, administrative in nature. The TS 
requirements involve Technical Review and Control and Audits. In 
accordance with the guidance provided in NRC Administrative Letter 
95-06, ``Relocation of Technical Specification Administrative 
Controls related to Quality Assurance,'' the proposed changes are an 
acceptable method for removing technical specification quality 
assurance requirements.
    The Independent Safety Review Function is being deleted because 
it is a redundant independent safety review to the existing 
independent review process being performed under the AmerGen/Exelon 
PORC.
    The remaining proposed changes are administrative in nature and 
have no affect on plant operation. The changes do not reduce the 
duties and responsibilities of the organizations performing the 
technical review, independent safety review and audit functions 
essential to ensuring the safe operation of the plant.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are administrative in nature. The proposed 
changes do not alter the physical design, safety limits, or safety 
analysis assumptions associated with the operation of the plant. 
Accordingly, the changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component to perform their safety 
function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes conform to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
guidance is presented in Administrative Letter 95-06, NUREG-1430[, 
``Standard Technical Specifications--Babcock and Wilcox Plants,''] 
and NUREG-1433[, ``Standard Technical Specifications--General 
Electric Plants, BWR/4.''] The relocation of these

[[Page 19109]]

administrative requirements and the deletion of a redundant 
independent safety review function will not reduce the 
[effectiveness of the] quality assurance commitments as accepted by 
the [Nuclear Regulatory Commission (NRC)], nor reduce administrative 
controls essential to the safe operation of the plant. Future 
changes to these administrative requirements will be performed in 
accordance with NRC regulation 10 CFR 50.54(a), consistent with the 
guidance identified above. Accordingly, the replacement of TS 
requirements by existing QATR requirements results in an 
[acceptable] level of regulatory control.
    Therefore, the proposed changes do not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Richard B. Ennis (Acting).

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

    Date of amendment request: November 13, 2007.
    Description of amendment request: The proposed amendment would 
delete the TMI Technical Specification (TS) 6.5, ``Review and Audit.'' 
Specifically, the proposed change would delete TS 6.5.1, ``Technical 
Review and Control,'' TS 6.5.2, ``Independent Safety Review Function,'' 
and TS 6.5.3, ``Audits'' which are currently being implemented by the 
Exelon/AmerGen Quality Assurance Topical Report. Additionally, the 
proposed amendment would correct typographical errors in the TMI 
Facility Operating License, the TS Table of Contents, and Figure 5-3 
while providing more legible versions of Figure 3.1-2a, Figure 3.5-1, 
Figure 3.5-2, and Figure 3.5-3. Further, the proposed amendment would 
update the description of the installed spent fuel pool storage 
locations. This request for approval of a license amendment was 
submitted concurrently with a similar request for the Oyster Creek 
Nuclear Generating Station (Oyster Creek). Due to some differences in 
the requested changes, the Oyster Creek amendment request will be 
noticed separately.
    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.
    No physical changes to the facilities, OCNGS [Oyster Creek] and 
TMI, will occur as a result of this proposed amendment. The proposed 
changes will not alter the physical design or operational procedures 
associated with any plant structure, system, or component.
    The proposed changes involve the deletion of several 
administrative requirements from the Technical Specifications (TS) 
that are now controlled under the Exelon/AmerGen Quality Assurance 
Topical Report (QATR) and several administrative procedures, AS-AA-
102, [Station Qualified Review Program], HU-AA-1212 (Independent 
Third Party Reviews), LS-AA-101 (License/TS changes), LS-AA-106 
(PORC), NO-AA-200-002 (Audits) and SY-AA-101-104 (Security Plan 
changes), and are, therefore, administrative in nature. The TS 
requirements involve Technical Review and Control and Audits. In 
accordance with the guidance provided in NRC Administrative Letter 
95-06, ``Relocation of Technical Specification Administrative 
Controls related to Quality Assurance,'' the proposed changes are an 
acceptable method for removing technical specification quality 
assurance requirements.
    The Independent Safety Review Function is being deleted because 
it is a redundant independent safety review to the existing 
independent review process being performed under the AmerGen/Exelon 
PORC.
    The remaining proposed changes are administrative in nature and 
have no affect on plant operation. The changes do not reduce the 
duties and responsibilities of the organizations performing the 
technical review, independent safety review and audit functions 
essential to ensuring the safe operation of the plant.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are administrative in nature. The proposed 
changes do not alter the physical design, safety limits, or safety 
analysis assumptions associated with the operation of the plant. 
Accordingly, the changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component to perform their safety 
function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes conform to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
guidance is presented in Administrative Letter 95-06, NUREG-1430[, 
``Standard Technical Specifications--Babcock and Wilcox Plants,''] 
and NUREG-1433[, ``Standard Technical Specifications--General 
Electric Plants, BWR/4.''] The relocation of these administrative 
requirements and the deletion of a redundant independent safety 
review function will not reduce the [effectiveness of the] quality 
assurance commitments as accepted by the [Nuclear Regulatory 
Commission (NRC)], nor reduce administrative controls essential to 
the safe operation of the plant. Future changes to these 
administrative requirements will be performed in accordance with NRC 
regulation 10 CFR 50.54(a), consistent with the guidance identified 
above. Accordingly, the replacement of TS requirements by existing 
QATR requirements results in an [acceptable] level of regulatory 
control.
    Therefore, the proposed changes do not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Richard B. Ennis (Acting).

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York

    Date of amendment request: February 28, 2008.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements related to the 
containment buffering agent used for pH control under post loss-of-
coolant accident (LOCA) conditions. Specifically, the proposal would 
approve the use of sodium tetraborate (STB) as the buffering agent 
instead of the currently approved compound, sodium hydroxide (NaOH). 
The reason for this change in buffering agents is to minimize the 
potential for an adverse chemical interaction between the NaOH and 
certain insulation materials in the containment that could degrade flow 
through the sump screens following certain design-basis accident 
scenarios such as a LOCA.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 19110]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because the 
containment buffering agent is not an initiator of any analyzed 
accident. The proposed change does not impact any failure modes that 
could lead to an accident.
    The proposed amendment does not involve a significant increase 
in the consequences of an accident previously evaluated. The 
buffering agent in containment is designed to buffer the acids 
expected to be produced after a LOCA and is credited in the 
radiological analysis for iodine retention. Utilizing STB as a 
buffering agent ensures the post-LOCA containment sump mixture will 
have a pH >= 7.0. The proposed change of replacing sodium hydroxide 
with STB results in the radiological consequences remaining within 
the limits of 10 CFR 50.67. There is no dose change with the pH 
above 7.0.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. STB is a passive component that is proposed to be used at 
IP3 as a buffering agent to increase the pH of the initially acidic 
post-LOCA containment water to a more neutral pH. Changing the 
proposed buffering agent from sodium hydroxide to STB does not 
constitute an accident initiator or create a new or different kind 
of accident previously analyzed. The operation of the Containment 
Spray System remains the same with the isolation of the sodium 
hydroxide to the eductors because the flow path of the spray remains 
constant through the eductors. The proposed amendment does not 
involve operation of any required systems, structures or components 
in a manner or configuration different from those previously 
recognized or evaluated. No new failure mechanisms will be 
introduced by the changes being requested.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment does not involve a significant reduction 
in a margin of safety. The proposed amendment of changing the 
buffering agent from sodium hydroxide to STB results in equivalent 
control of maintaining sump pH at 7.0 or greater, thereby 
controlling containment atmosphere iodine and ensuring the 
radiological consequences of a LOCA are within regulatory limits. 
The change of buffering agent from NaOH to STB also reduces the 
amount of sodium aluminum silicate precipitate thereby reducing the 
overall amount of precipitate that may be formed in a postulated 
LOCA. The buffer change would minimize the potential chemical 
effects and should enhance the ability of the emergency core cooling 
system to perform the post-accident mitigating functions.
    Therefore, the proposed amendment does not involve a significant 
reduction in [a] margin of safety.

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

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: February 25, 2008.
    Description of amendment request: The proposed amendment would 
revise NMP1 Technical Specification (TS) 3/4.4.4, ``Emergency 
Ventilation System,'' to remove the operability and surveillance 
requirements for the 10,000 watt heater located in the common supply 
inlet air duct for the Reactor Building Emergency Ventilation System 
(RBEVS). The proposed amendment would also revise TS 3/4.4.5, ``Control 
Room Air Treatment System,'' to reduce the 10-hour duration monthly 
system operational surveillance test requirement to a 15-minute run 
surveillance test requirement.
    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 RBEVS and the CRAT [Control Room Air Treatment] System do 
not involve any initiators or precursors to an accident previously 
evaluated as the systems perform a mitigative function in response 
to an accident. Failure of the systems would result in the inability 
to perform their mitigative function but would not increase the 
probability of an accident previously evaluated. The RBEVS is 
designed to limit the release of radioactive gases to the 
environment such that resulting doses will be less than the 
guideline values of 10 CFR 50.67, ``Accident Source Term.'' The CRAT 
System is designed to minimize the amount of radioactivity or other 
gases from entering the control room in the event of an accident. 
Both the RBEVS and the CRAT System charcoal filter materials are 
tested in accordance with American Society for Testing and Materials 
(ASTM) D3803-1989, ``Standard Test Method for Nuclear-Grade 
Activated Carbon'' at a test temperature of 30[deg] C [degrees 
Celsius] (86[deg]F [degrees Fahrenheit]) while maintaining a 
relative humidity (RH) value of 95%. The testing method assures the 
ability of the charcoal filters to perform their intended function 
with or without the humidity control function provided by the 10 kW 
[kilowatt] heater. The filter efficiency values required by the TS 
test criteria provide a safety factor of 2, consistent with the 
recommendations of GL [Generic Letter] 99-02.
    The previous NMP1 adoption of the more stringent ASTM D3803-1989 
charcoal testing parameters resulted in the elimination for humidity 
control of inlet air in both the RBEVS and the CRAT system. 
Therefore, the need for a 10 hour duration monthly system 
operational surveillance test is no longer necessary to demonstrate 
reliability and proper function of the systems.
    Therefore, these proposed changes do not alter the results of 
the accident dose consequence analyses and do not involve a 
significant increase in 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 amendment of removing the RBEVS 10 kW common supply 
air inlet heater requirements and reducing the duration of the 
monthly system operational surveillance requirements from 10 hours 
to 15 minutes for both the RBEVS and the CRAT System will not 
involve placing the system in a new configuration or operating the 
system in a different manner that could result in a new or different 
kind of accident. Testing of the charcoal filter materials in 
accordance with ASTM D3803-1989 standard at a test temperature of 
30[deg] C (86[deg] F) while maintaining a relative humidity 95% will 
continue to assure the ability of the system's charcoal filters to 
perform its intended function under potential higher inlet air RH 
values.
    The previous NMP1 adoption of the more stringent ASTM D3803-1989 
charcoal testing parameters resulted in the elimination of the need 
for humidity control of inlet air in both the RBEVS and the CRAT 
system. Therefore, a 10 hour duration monthly system operational 
surveillance test is no longer necessary to demonstrate reliability 
and proper function of the systems.

[[Page 19111]]

    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any [accident] previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment will not adversely affect the performance 
characteristics of either the RBEVS or the CRAT System, and will not 
affect the ability of either system to perform its intended 
function.
    Testing of the charcoal filter materials in accordance with ASTM 
D3803-1989 and the test parameters required by the TS assures the 
ability of the charcoal filters to perform their intended function 
with or without the humidity control. The filter efficiency values 
required by the TS test criteria provide a safety factor of 2, 
consistent with the recommendations of GL 99-02.
    The previous NMP1 adoption of the more stringent ASTM D3803-1989 
charcoal testing parameters resulted in the elimination of the need 
for humidity control of inlet air in both the RBEVS and the CRAT 
system. Therefore, a 10 hour duration monthly system operational 
surveillance test is no longer necessary to demonstrate reliability 
and proper function of the systems.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: February 8, 2008.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.6.6, ``Reactor Coolant System (RCS) 
Pressure and Temperature Limits Report (PTLR),'' to update the method 
used to develop the RCS heatup and cooldown and Low Temperature Over 
Pressure (LTOP) limits utilizing current NRC approved methodology.
    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.
    Integrity of the reactor vessel is integral to plant safety. It 
provides containment and continuity for the reactor core, and as 
part of the reactor coolant system acts as one of the three fission 
product barriers to the environment. The purpose of the heatup and 
cooldown limit curves and LTOP setpoints is to ensure vessel 
integrity through the spectrum of operating modes. Operating within 
those limits ensures that brittle failure of the vessel material 
does not occur due to the thermal and pressure stresses the vessel 
is subjected to during operation. During power operation, the 
effects of neutron radiation tend to change the characteristics of 
the vessel material making it more brittle. To compensate for this 
the operating limits must be periodically adjusted. The methodology 
being proposed in this submittal is designed to ensure vessel 
integrity, is analytically sound, and has been reviewed and approved 
by the NRC. 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.
    Due to its thickness and material properties, the reactor vessel 
is the limiting component for brittle fracture in the reactor 
coolant system. The proposed methodology appropriately limits the 
operating parameters to preclude the possibility of vessel failure. 
No new failure mechanisms or accident precursors are introduced as a 
result of this proposed change. Therefore, this 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 methodology in WCAP-14040-A, Methodology Used to 
Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup 
and Cooldown Limit Curves, Revision 4, contains appropriate margin 
and has been reviewed and approved by the NRC. Since the new 
methodology for developing heatup and cooldown curves will produce 
less restrictive curves, use of the existing methodology for LTOP 
setpoints will continue to provide adequate margin to the [Title 10 
of the Code of Federal Regulations, Part 50] Appendix G limits. 
Therefore, this proposed change does not involve a significant 
reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

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.

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

    Date of amendment request: February 25, 2008.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) 5.5.9, ``Steam Generator (SG) 
Program,'' and TS 5.6.9, ``Steam Generator Tube Inspection Report.'' 
For TS 5.5.9, the amendment would replace the existing alternate repair 
criteria in the provisions for SG tube repair criteria during Braidwood 
Station (Braidwood), Unit 2, refueling outage 13 and the subsequent 
operating cycle. For TS 5.6.9, three new reporting requirements are 
proposed to be added to the existing seven requirements. The proposed 
changes only affect Braidwood, Unit 2; however, this is docketed for 
Braidwood, Units 1 and 2, because the TS are common to both units.
    Date of publication of individual notice in Federal Register: March 
11, 2008.
    Expiration date of individual notice: April 11, 2008 (public 
comment); May 11, 2008 (hearing requests).

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

    Date of amendment request: March 7, 2008.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 4.3-1, ``Reactor Trip System 
Instrumentation

[[Page 19112]]

Surveillance Requirements,'' to require the initial plateau curves to 
be measured within 24 hours after attaining 100 percent steady-state 
power. Currently, initial plateau curves are required to be taken 
within 24 hours of entry into Mode 2.
    Date of publication of individual notice in Federal Register: March 
19, 2008 (FR 72 14850).
    Expiration date of individual notice: April 18, 2008 (Public 
comment) and May 19, 2008 (Hearing requests).

Notice of Issuance of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, 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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of application for amendments: October 17, 2007, as 
supplemented by letter dated December 13, 2007.
    Brief description of amendments: These amendments modify Technical 
Specification requirements for inoperable snubbers by adding Limiting 
Condition for Operation (LCO) 3.0.8. This operating license improvement 
was made available by the Nuclear Regulatory Commission on May 4, 2005 
(70 FR 23252) as part of the consolidated line item improvement 
process. In addition, the amendments correct an omission to Amendment 
No. 282 and 259, issued on September 27, 2007, that adopted Technical 
Specification Task Force Traveler 427 by including a reference to LCO 
3.0.9 in LCO 3.0.1.
    Date of issuance: March 24, 2008.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: Unit 1--285, Unit 2--262.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65362) The letter dated December 13, 2007, 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 these amendments is 
contained in a Safety Evaluation dated March 24, 2008.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: March 19, 2007, as supplemented 
by letter dated August 16, 2007.
    Brief description of amendment: The amendment revised the Technical 
Specification 3.8.1 entitled ``AC [Alternating Current] Sources-
Operating'' to change the minimum Emergency Diesel Generator output 
voltage acceptance criterion from 3740 to 3873 volts. Specifically, the 
proposed change revised the Surveillance Requirements 3.8.1.2, 3.8.1.7, 
3.8.1.10, 3.8.1.11, 3.8.1.14, and 3.8.1.17.
    Date of issuance: March 17, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 178.
    Facility Operating License No. NPF-43: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20379).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2008.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket 
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 
2, Rock Island County, Illinois

    Date of application for amendments: April 12, 2007.
    Brief description of amendments: The amendments revise technical 
specification requirements related to control room envelope 
habitability in accordance with Technical Specification Task Force 
Traveler TSTF-448, Revision 3, ``Control Room Habitability.''
    Date of issuance: March 20, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of the date of issuance.
    Amendment Nos.: 226/218 and 238/233.
    Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and 
DPR-30: The amendments revised the Technical Specifications and 
License.
    Date of initial notice in Federal Register: June 5, 2007 (72 FR 
31100) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 20, 2008.
    No significant hazards consideration comments received: No.

[[Page 19113]]

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

    Date of application for amendments: November 20, 2007.
    Brief description of amendments: The amendments revised the values 
of the safety limit minimum critical power ratio (SLMCPR) in Technical 
Specification (TS) Section 2.1.1, ``Reactor Core SLs.'' Specifically, 
the amendments deleted the Quad Cities Nuclear Power Station (QCNPS), 
Unit 2 fuel-specific SLMCPR requirements for Global Nuclear Fuel GE14 
fuel and consolidated QCNPS SLMCPR requirements into a bounding dual-
unit requirement.
    Date of issuance: February 28, 2008.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the next refueling outage for QCNPS, Unit 2 
(Q2R19), which is scheduled to start in March 2008.
    Amendment Nos.: 237/232.
    Renewed Facility Operating License Nos. DPR-29 and DPR-30: The 
amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71712) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 28, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 9, 2007, as 
supplemented by letters dated August 8, August 23, September 13, 2007, 
and January 25, 2008.
    Brief description of amendment: The amendment will address Generic 
Safety Issue 191 ``Assessment of Debris Accumulation on PWR Sump 
Performance,'' by implementing Technical Specification (TS) changes 
that reflect the use of a new recirculation spray system pump start 
signal due to a modification to the containment sump screens and 
replace the use of LOCTIC with the Modular Accident Analysis Program-
Design Basis Accident calculation methodology to calculate containment 
pressure, temperature, and condensation rates for input to the SWNAUA 
code, which ultimately changes the aerosol removal coefficients used in 
dose consequence analysis.
    Date of issuance: March 11, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented prior to the first entry into Mode 4 coming out of 2R13, 
which begins April 2008.
    Amendment No: 164.
    Facility Operating License No. NPF-73: The amendment revised the 
License and TS.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20383). The supplements dated August 8, August 23, September 13, 2007, 
and January 25, 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 March 11, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 14, 2007.
    Brief description of amendment: This amendment consists of changes 
to the Technical Specifications in response to your application dated 
September 14, 2007, which requested revision to TS 3.3.2.1, ``Control 
Rod Block Instrumentation,'' Table 3.3.2.1-1, ``Control Rod Block 
Instrumentation,'' to modify a footnote such that a new Banked Position 
Withdrawal Sequence (BPWS) shutdown sequence could be utilized.
    Date of issuance: March 20, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 268.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications and Facility Operating License.
    Date of initial notice in Federal Register: December 4, 2007 (72 FR 
68216).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 20, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 2, 2007, as 
supplemented by letters dated February 8, and March 11, 2008.
    Brief description of amendments: The amendments revised Technical 
Specification 3.5.4, ``Refueling Water Storage Tank (RWST),'' and 
Surveillance Requirement 3.5.4.2, to increase the minimum required 
borated water volume from ``>= 400,000 gallons (81.5% indicated 
level)'' to ``>= 455,300 gallons.''
    Date of issuance: March 26, 2008.
    Effective date: As of its date of issuance and shall be implemented 
prior to Mode 4 entry following refueling outage 2R14.
    Amendment Nos.: Unit 1--199; Unit 2--200.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: December 31, 2007 (72 
FR 74361). The supplemental letters dated February 8, and March 11, 
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 26, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 17, 2007, as 
supplemented by letter dated December 13, 2007.
    Brief description of amendment: The amendment revises Technical 
Specifications requirements for inoperable snubbers by adding Limiting 
Condition for Operation 3.0.8. This operating license improvement was 
made available by the Nuclear Regulatory Commission on May 4, 2005 (70 
FR 23252) as part of the consolidated line item improvement process.
    Date of issuance: March 19, 2008.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 104.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65371) The letter dated December 13, 2007, provided additional 
information that clarified the application, did not

[[Page 19114]]

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 March 19, 2008.
    No significant hazards consideration comments received: No.

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 application for amendments: February 8, 2007, as 
supplemented by letters dated July 24 and November 15, 2007, and 
February 19, 2008.
    Brief description of amendments: The amendments incorporate changes 
which (1) revised Technical Specifications (TS) Surveillance 
Requirement (SR) 3.3.7.3.a to lower the allowable value for dropout and 
raise the allowable value for pickup of the degraded voltage function, 
and (2) revised TS SR 3.8.1 to lower the diesel generator minimum 
output voltage due to lower settings for the degraded voltage function.
    Date of issuance: March 18, 2008.
    Effective date: as of its date of issuance, to be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2--216; Unit 3--208.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 27, 2007 (72 FR 
14307). The supplemental letters dated July 24 and November 15, 2007, 
and February 19, 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 18, 2008.
    No significant hazards consideration comments received: No.

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 application for amendments: March 30, 2007, as supplemented 
by letters dated November 5, 2007, and January 15 and February 19, 
2008.
    Brief description of amendments: The amendments revised Technical 
Specifications Surveillance Requirement 3.3.7.3.b, ``Loss of Voltage 
Function,'' to a narrower voltage band and lower operating time for 
channel calibration testing, by replacing the undervoltage relays with 
the reset time significantly lower.
    Date of issuance: March 25, 2008.
    Effective date: As of the date of its issuance and shall be 
implemented in the next refueling outage or unit outage of sufficient 
duration, whichever occurs first.
    Amendment Nos.: Unit 2--217; Unit 3--209.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20385). The supplemental letters dated November 5, 2007, and January 15 
and February 19, 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 25, 2008.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 21, 2007, as supplemented by letter 
dated November 26, 2007.
    Brief description of amendments: The amendments revised the 
Technical Specifications Surveillance Requirement 4.5.2.d for the 
Emergency Core Cooling System sumps for consistency with the new sump 
design and configuration.
    Date of issuance: March 25, 2008.
    Effective date: As of its date of issuance and shall be implemented 
after completion of the corrective actions and modifications for 
resolution of Generic Safety Issue 191, ``Assessment of Debris 
Accumulation on PWR Sump Pump Performance.''
    Amendment Nos.: Unit 1--183; Unit 2--170.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 31, 2007 (72 FR 
41789). The supplemental letter dated November 26, 2007, 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 25, 2008.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 14, 2007, as supplemented by 
letters dated September 12, October 16, and December 14 (two letters), 
2007, and January 18, 2008.
    Brief description of amendment: The amendment authorizes (1) the 
replacement of the main steam isolation valves (MSIVs) and main 
feedwater isolation valves (MFIVs) and (2) the use of Figures B 3.7.2-1 
(MSIVs) and 3.7.3-1 (MFIVs) as the limiting closure times for these 
valves to demonstrate that these valves meet the limiting conditions 
for operation with respect to the valve closure time. The remaining 
amendment requests in the application that have not yet been addressed 
by the NRC are the proposed (1) addition of main feedwater regulating 
valves and bypass valves to TS 3.7.3, ``Main Feedwater Isolation 
Valves,'' and (2) modification of the main steam and feedwater 
isolation system (MSFIS) controls. These requests will be addressed in 
future letters to the licensee.
    Date of issuance: March 21, 2008.
    Effective date: As of its date of issuance and shall be implemented 
before entry into Mode 3 in the restart from Refueling Outage 16, which 
is to be conducted in the spring