[Federal Register Volume 75, Number 75 (Tuesday, April 20, 2010)]
[Notices]
[Pages 20627-20644]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-8744]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0156]


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) 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 25, 2010 to April 7, 2010. The last 
biweekly notice was published on April 6, 2010 (75 FR 17439).

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 Title 10 of the Code of Federal 
Regulations (10 CFR), Section 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

[[Page 20628]]

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, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area O1F21, 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 by the above date, 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 identify the specific contentions which the requestor/petitioner 
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 
requestor/petitioner 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 requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner 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 requestor/petitioner to relief. A requestor/
petitioner 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, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
[email protected], or by telephone at (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 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    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. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.

[[Page 20629]]

    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through EIE, users will be required to install a Web 
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser 
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant 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 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing 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 
E-Filing 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 using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at  http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
[email protected], or by a toll-free call at (866) 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, 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. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHDProceeding/home.asp., unless excluded pursuant to 
an order of the Commission, or the presiding officer. Participants are 
requested not to include personal privacy information, such as social 
security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, 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 O1F21, 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. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].

Arizona Public Service Company, 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: November 30, 2009.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.3.5, ``Engineered Safety Features 
Actuation System Instrumentation,'' Table 3.3.5-1, to raise the 
refueling water tank (RWT) low level allowable values for the 
recirculation actuation signal (RAS); raise the minimum required RWT 
volume shown in TS Figure 3.5.5-1; and implement a time-critical 
operator action to close the RWT isolation valves, including 
consideration of a potentially more limiting single failure of a low-
pressure safety injection pump to automatically stop, as designed, on 
an RAS.
    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 RWT is a passive component of the Chemical and Volume 
Control System (CVCS) that supports ECCS [emergency core cooling 
system] and CSS [containment spray system] operation to mitigate the 
consequences of an accident. A[n] RAS is an active component of the 
Engineered Safety Features Actuation System (ESFAS) that actuates 
safety equipment to mitigate the consequences of a LOCA [loss-of-
coolant accident]. Neither of these components initiates an accident 
previously evaluated. The RWT isolation valves are also components 
of the CVCS; however, their closure was not previously credited for 
RWT isolation following a[n] RAS. The proposed amendment will credit 
closure of these valves following a[n] RAS to preclude the potential 
for air entrainment in the ECCS and CS [containment spray] pump 
suction piping for any LOCA scenario. The required isolation is 
being performed as a time critical

[[Page 20630]]

operator action, which is consistent with ANSI/ANS-58.8-1984 
[American National Standards Institute/American Nuclear Society 
Standard 58.8-1984], Time Response Design Criteria for Safety-
Related Operator Actions, 1984 guidance. Although the change in the 
closure requirement and the operator action could introduce 
additional potential malfunctions, these malfunctions have been 
evaluated and found not to initiate or have a significant adverse 
affect on the mitigation or consequences of any accident previously 
evaluated.
    The proposed changes do not alter or prevent the ability of 
structures, systems or components to perform their intended function 
to mitigate the consequences of an initiating event within the 
assumed acceptance limits. The proposed changes will ensure 
continued performance of the ECCS and CS pumps following a LOCA by 
precluding the potential for air entrainment in the pump suction 
piping from the RWT after a[n] RAS.
    The effect of the proposed changes to the RAS Allowable Values 
and RWT minimum required level on the RWT structural design, 
containment post-LOCA flood level, post-LOCA boron precipitation, 
and containment sump pH remain within the limits assumed in the 
design and accident analyses. The proposed license amendment does 
not affect the source term, containment isolation, or radiological 
release assumptions used in evaluating the radiological consequences 
of an accident previously evaluated. Further, the proposed changes 
do not increase the types or amounts of radioactive effluent that 
may be released offsite. The proposed license amendment is 
consistent with these analyses' assumptions and resultant 
consequences.
    The proposed amendment also recognizes and evaluates a different 
single failure associated with the RWT drain down following a LOCA 
than previously evaluated. It was determined this failure was of low 
probability and did not adversely affect any previous bounding 
analysis or the capability of the associated systems to perform 
their design functions.
    Therefore, the proposed license amendment 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 license amendment does not involve or add any new 
or different components to the plant and does not change any 
accident initiators.
    The proposed changes to the RAS Allowable Values and RWT minimum 
required level will not change the design function of the RWT to 
support ECCS and CSS operation following a LOCA. However, the 
closure of the RWT isolation valves following a LOCA was not 
previously credited. As a result, the credited RWT isolation valve 
design function has been changed, and closure of these valves is now 
credited to preclude the possibility of air entrainment in the ECCS 
and CS pump suction piping for any LOCA scenarios. The credited 
isolation is being performed as a time critical operator action, 
which is consistent with ANSI/ANS 58.8 guidance. Although changes to 
the valve closure requirement and the operator action introduce 
additional potential malfunctions, these malfunctions have been 
evaluated and found not to create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendment recognizes and evaluates a different 
single failure associated with the RWT drain down following a LOCA 
than previously evaluated. It was determined that this failure was 
of low probability and did not adversely affect any previous 
bounding analysis or create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed license amendment does not alter the manner in 
which safety limits, limiting safety system settings, or limiting 
conditions for operation are determined or implemented. The safety 
analysis acceptance criteria are not affected by this amendment. The 
proposed changes in the credited design function of the RWT 
isolation valves, along with the change in the RAS Allowable Value 
and RWT minimum required levels, continue to ensure sufficient RWT 
water volume to enable the ECCS and CSS to satisfy required design 
functions for all postulated LOCA break sizes. Therefore, these 
changes do not impact the results of safety analyses.
    The proposed changes to the RAS Allowable Values and minimum 
required RWT level include appropriate instrument uncertainties and 
are based on conservative analyses for establishing the required RWT 
volumes. The proposed amendment will not result in plant operation 
in a configuration outside of the design basis.
    The proposed amendment recognizes and evaluates a different 
single failure associated with the RWT drain down following a LOCA 
than previously evaluated. It was determined this failure was of low 
probability and did not adversely affect any previous bounding 
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 
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 no significant hazards consideration.
    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: Michael T. Markley.

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

    Date of amendment request: January 29, 2010.
    Description of amendment request: The amendment would modify the 
existing Note within Technical Specification 3.4.10, ``Pressurizer 
Safety Valves [PSVs],'' which covers operation in the applicable 
portions of Mode 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No.
    The proposed change, revising an existing NOTE within Technical 
Specification 3.4.10 to allow the PSVs lift settings to be outside 
LCO [Limiting Condition for Operation] values, as a result of 
temperature related drift, while the Unit is in applicable portions 
of Mode 3 for periods up to 36 hours, does not change the design 
function or operation of the PSVs and it does not change the way the 
PSVs are maintained, tested, or inspected. In addition the proposed 
change does not change any of the evaluated accidents in our Updated 
Final Safety Analysis Report, does not change PSV lift settings, or 
impact the ability of the PSVs to perform their safety function 
during evaluated accidents.
    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?
    No.
    The proposed change, revising an existing NOTE within Technical 
Specification 3.4.10 to allow the PSVs lift settings to be outside 
LCO values, as a result of temperature related drift, while the Unit 
is in applicable portions of Mode 3 for periods up to 36 hours, does 
not change the PSVs design function to maintain RCS [reactor coolant 
system] pressure below the RCS pressure Safety Limit of 2750 psia 
during design basis accidents nor does it affect the PSVs ability to 
perform this design function. The proposed change does not require 
any modification to the plant or change equipment operation or 
testing. It also does not create any credible new failure 
mechanisms, malfunctions, or accident initiators that would cause an 
accident not previously considered.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

[[Page 20631]]

    No.
    The proposed change, revising an existing NOTE within Technical 
Specification 3.4.10 to allow the PSVs lift settings to be outside 
LCO values, as a result of temperature related drift, while the Unit 
is in applicable portions of Mode 3 for periods up to 36 hours, does 
not involve a significant reduction in the margin of safety in 
maintaining RCS pressure below Safety Limits of 2750 psia during 
design basis accidents. The analysis conducted in support of this 
proposed change evaluated the ability of the PSVs to maintain an 
adequate safety margin when required in applicable Mode 3 conditions 
despite the identified temperature related lift setting drift. The 
analysis identified that there were no credible design accident 
scenarios, when in the applicable Mode 3 conditions, that challenged 
the PSVs to respond in order to maintain an adequate safety margin 
to the reactor coolant Safety Limit of 2750 psia.
    Therefore the proposed change does not involve a significant 
reduction in the margin of safety of maintaining RCS pressure below 
the RCS pressure Safety Limit.

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

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

    Date of amendment request: January 4, 2010.
    Description of amendment request: The proposed amendment would 
revise the Core Spray flow requirement in Technical Specifications 
Surveillance Requirements 3.5.1.8 and 3.5.2.6 from 6,350 to 5,725 
gallons per minute consistent with the flow assumed in the Emergency 
Core Cooling System (ECCS) safety analyses.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The minimum performance requirements of the low pressure 
Emergency Core Cooling System (ECCS) pumps, including the Core Spray 
pumps, are determined through application of the 10 CFR 50, Appendix 
K methodology to ensure the criteria of 10 CFR 50.46 are satisfied. 
The surveillance testing of the Core Spray pumps is performed 
periodically in accordance with the ASME Code, Section XI verifies 
that two Core Spray pumps in parallel operation within a single 
division develop sufficient discharge pressure at the Technical 
Specification required flow to overcome the elevation head pressure 
between the pump suction and the vessel discharge, the piping 
friction losses, and TS SR specified Reactor Pressure Vessel 
pressure. The acceptance criteria necessary to satisfy the revised 
TS SRs would be established in the plant design basis in the form of 
the minimum required pump performance defined for a range of flow 
about the specified TS SR flow. Detroit Edison intends to continue 
TS SR and IST pump testing at the current IST pump baseline flow and 
establish compliance with the TS SR by comparing the measured 
performance against the design minimum pump curve. In this manner, 
the minimum actual delivered divisional Core Spray pump performance 
is assured to meet or exceed that required by the Appendix K safety 
analyses. These performance requirements are unchanged and are met 
by the proposed change.
    The bases for the core spray flow requirements in the Technical 
Specifications Surveillance Requirements are unchanged. The 
requirements are selected based on the flow values assumed and used 
in the current ECCS safety analyses. The value proposed for core 
spray divisional (2 pump) flow is consistent with the inputs used 
for ECCS safety analyses performed for the current licensed power 
level.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 revises the Technical Specification 
Surveillance Requirements for Core Spray flow to be consistent with 
the accident analysis. No physical changes are being made to the 
installed core spray system. The proposed surveillance requirements 
are consistent with those used in the accident analyses which 
analyze the effect of Core Spray system performance for the accident 
conditions for which the system is designed to respond. No new or 
different accident scenarios are created by this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The Core Spray system has historically been capable of meeting 
the Core Spray Technical Specification Surveillance Requirements. 
However, correction of non-conservative errors in the system 
hydraulic calculation and the identification of a non-conservative 
bias in the test flow instrument calibration have eroded the test 
margin such that it is possible that the Technical Specification 
Surveillance Requirements may not be satisfied for some 
surveillances and at the same time maintain a relatively large 
margin compared to the minimum performance assumed in the ECCS 
safety analyses. These non-conservative errors or biases have always 
existed, but have not always been specifically accounted for in the 
surveillance testing acceptance criteria. Since there is no change 
in the Technical Specification bases associated with the requested 
change, there is no real change in the margin provided in the system 
design or analyses. The proposed change makes the margin between the 
current Core Spray Technical Specification Surveillance Requirements 
and the performance assumed in the plant safety analyses available 
as a design and test margin. The minimum required performance 
necessary to satisfy the Core Spray Technical Specification 
Surveillance Requirements will be established in the plant design 
basis with the minimum required pump performance adjusted upward as 
necessary to account for instrument uncertainty and bias as well as 
differences between assumed accident and actual test operating 
conditions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: November 23, 2009, as supplemented by 
letter dated March 18, 2010.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) requirements for testing of 
the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Safety/Relief 
Valves (SRVs) by replacing the current requirement to manually actuate 
each SRV during plant startup with a requirement to verify that each 
valve is capable of being opened. The proposed amendment would change 
both TS Surveillance Requirements (SRs) 3.4.3.2 and 3.5.1.13 to verify 
that each required valve ``is capable of being opened.'' The current 
Frequency for both TS SRs is ``24 months on a STAGGERED TEST BASIS for 
each valve solenoid''; this

[[Page 20632]]

would be changed to state, ``In accordance with the Inservice Testing 
Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not modify the method of demonstrating 
the Operability of the Safety/Relief Valves (SRVs) in both the 
safety and relief modes of operation. As currently stated in the 
Bases ``...valve OPERABILITY and the setpoints for overpressure 
protection are verified, per ASME Code requirements, prior to valve 
installation.'' The proposed change does modify the method for 
demonstrating the proper mechanical functioning of the SRVs and that 
the valves and discharge lines are free of obstructions. The SRVs 
are required to function in the safety mode to prevent 
overpressurization of the reactor vessel and reactor coolant system 
pressure boundary during various analyzed transients, including Main 
Steam Isolation Valve closure. SRVs associated with the Automatic 
Depressurization System are also required to function in the relief 
mode to reduce reactor pressure to permit injection by low pressure 
Emergency Core Cooling System (ECCS) pumps during certain reactor 
coolant pipe break accidents. The current testing method 
demonstrates the proper mechanical functioning of the SRVs in both 
modes through manual actuation of the SRVs. The proposed new testing 
method demonstrates both Operability and proper mechanical 
functioning using a series of overlapping tests that demonstrate 
proper functioning of the SRV stages and supporting control 
components. This proposed testing method results in acceptable 
demonstration of the SRV functions in both the safety and relief 
modes, and therefore provides assurance that the probability of SRV 
failure will not increase. None of the accident safety analyses is 
affected by the requested Technical Specifications (TS) changes. 
Therefore, the consequences of accidents mitigated by the SRVs will 
not increase.
    Certain SRV malfunctions are included in the FSAR [final safety 
analysis report] safety analyses. Specifically, the plant safety 
analyses include the inadvertent opening of an SRV and a stuck open 
SRV. By not actuating the SRVs during plant operation for testing 
and thus reducing the incidence of pilot stage leakage of the SRVs, 
the proposed testing eliminates a contributor to these events.
    Based on these considerations, the proposed test method does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change modifies the method of testing of the SRVs, 
but does not alter the functions or functional capabilities of the 
SRVs. Testing under the proposed method is performed in offsite test 
facilities or in the plant during outage periods when the SRV 
functions are not required. Existing analyses address events 
involving an SRV inadvertently opening or failing to reclose. 
Analyses also address the likelihood and consequences of failure of 
one or more SRVs to open. The proposed change does not introduce any 
new failure mode, and therefore, does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    Overpressure protection of the reactor coolant pressure boundary 
is based on the SRV setpoints and total relief capacity. Setpoint is 
verified at an offsite testing facility; this requirement is not 
altered by the proposed change. Relief capacity of each SRV is 
determined by valve geometry, which is also not altered by the test 
methods. The margin of safety in the Loss of Coolant Accident 
analysis due to operation of the Automatic Depressurization System 
is also based on total relief capacity of the associated SRVs. The 
proposed change in surveillance test methods demonstrates the 
operability of the SRVs, but does not alter the critical parameters 
that affect the margin of safety in analyses involving the SRV 
functions. Therefore, the proposed change does not involve a 
significant reduction in any 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: Nancy L. Salgado.

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

    Date of amendment request: February 22, 2010.
    Description of amendment request: The proposed amendment will allow 
implementation of leak-before-break (LBB) on the Waterford Steam 
Electric Station, Unit 3 (Waterford 3) pressurizer surge line. The 
licensee will be replacing the two Waterford 3 steam generators (SGs) 
during the forthcoming spring 2011 refueling outage. Based on design 
changes in the replacement SGs, piping systems will require rerouting 
in the SG cavity area. Due to the existing dynamic piping protection 
associated with the pressurizer surge line, rerouting of the 
replacement SG blowdown line cannot be effectively performed without 
the elimination of dynamic protection for the pressurizer surge line.
    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 uses an approved leak-before-break (LBB) 
fracture mechanics methodology, in accordance with 10CFR50 [Title 10 
of the Code of Federal Regulations, Part 50], Appendix A, General 
Design Criterion (GDC) 4 to demonstrate that the probability of 
fluid system rupture for these lines attached to the Reactor Coolant 
System (RCS) is extremely low under conditions associated with the 
design basis for the piping. The proposed change does not adversely 
affect accident initiators or precursors nor significantly alter the 
design assumptions, conditions, and configuration of the facility or 
the manner in which the plant is operated and maintained. Overall 
protection system performance will remain within the bounds of the 
previously performed accident analyses. The design of the protection 
systems will be unaffected. The Reactor Protection System (RPS) and 
Emergency Core Cooling System (ECCS) will continue to function in a 
manner consistent with the plant design basis. All design, material, 
and construction standards that were applicable prior to the request 
are maintained. There will be no change to normal plant operating 
parameters or accident mitigation performance. The proposed 
amendment will not alter any assumptions or change any mitigation 
actions in the radiological consequence evaluations in the FSAR 
[Final Safety Analysis Report].
    Therefore, this 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 does not create the possibility of a new or 
different kind of accident, since it provides an NRC acceptable 
alternate means for demonstrating that the probability of a fluid 
system rupture is extremely small. There are no changes in the 
methods by which any safety-related plant

[[Page 20633]]

system performs its safety function. No new accident scenarios, 
transient precursors, failure mechanisms, or limiting single 
failures are introduced as a result of this amendment. There will be 
no adverse effect or challenges imposed on any safety-related system 
as a result of this amendment. LBB methodology per GDC-4 still 
requires that ECCS, containment, and equipment qualification (EQ) 
requirements be maintained consistent with the original postulated 
accident assumptions. Only protection from dynamic effects is 
modified.
    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 changes apply conservative approved analytical 
methods to demonstrate that the probability of a fluid system 
rupture is very low. This analysis retains substantial margins to 
assure that pipe rupture is extremely low and justifies differences 
in protection from dynamic effects with these extremely low 
probability ruptures. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions. For overall ECCS, 
containment, and EQ requirements, there will be no changes to the 
assumed margins.
    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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: February 22, 2010.
    Description of amendment request: The proposed amendment would add 
valve SI-4052A (Reactor Coolant Loop (RCL) 2 Shutdown Cooling (SDC) 
suction inside containment bypass isolation) and valve SI-4052B (RCL 1 
SDC suction inside containment bypass isolation) to Technical 
Specification (TS) Table 3.4-1, ``Reactor Coolant System Pressure 
Isolation Valves.'' The purpose of this line is to equalize the SDC 
system pressure down stream of valve SI-405A (RCL 2 SDC suction inside 
containment isolation) and valve SI-405B (RCL 1 SDC suction inside 
containment isolation) in order to minimize the pressure transient in 
the system when valves SI-405A(B) are opened.
    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 addition of the bypass fill line will decrease the 
likelihood of a pressure transient in the Shutdown Cooling System 
suction piping which increases the reliability of the Shutdown 
Cooling System. Once this change is installed valves SI-405A(B) and 
SI-4052A(B) become parallel inside containment isolation valves in 
the shutdown cooling system suction lines. The configuration of SI-
405A(B) and SI-4052A(B) includes interlocks such that these valves 
cannot be inadvertently opened with the RCS [reactor coolant system] 
above the design pressure of the shutdown cooling system. This 
change does not affect the capability of these valves to isolate the 
RCS from SDC. Therefore, there is no credible mechanism by which 
this change can introduce an inter-system LOCA [loss-of-coolant 
accident] (ISLOCA) different than previously evaluated in the UFSAR 
[Updated Final Safety Analysis Report]. These features are, 
discussed in FSAR [Final Safety Analysis Report] section 7.6.1.1.2.
    Therefore, this 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.
    Once this change is installed valves SI-405A(B) and SI-4052A(B) 
become parallel inside containment isolation valves in the shutdown 
cooling system suction lines. SI-4052A(B) and its associated lines 
and valves are designed to the same requirements as SI-405A(B) and 
its associated lines. The previously evaluated SI-405A(B) failure 
modes bound those failure modes possible by SI-4052A(B). Thus, no 
failure of SI-4052A(B) exists that would be different or more severe 
than SI-405A(B),
    This 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 amendment adds SI-4052A(B) to Technical 
Specification Table 3.4-1. The change also adds an allowed leakage 
limit to SI-4052A(B) consistent with NUREG-1432 guidance.
    Since the SI-4052A(B) leakage limit is commensurate with the 
valve size, this does not represent 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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: February 22, 2010.
    Description of amendment request: Entergy Operations, Inc. (the 
licensee), will be replacing the two Waterford Steam Electric Station, 
Unit 3 (Waterford 3) steam generators (SGs) during the 17th refueling 
outage which will commence in the spring of 2011. The existing 
Waterford 3 SG program under Technical Specification (TS) 6.5.9 
contains an alternate repair criterion for SG tube inspections that is 
no longer applicable to the replacement SGs. The proposed amendment 
will modify TS 6.5.9, ``Steam Generator (SG) Program,'' and TS 6.9.1.5, 
``Steam Generator Tube Inspection Report,'' to eliminate currently 
allowed SG tube alternate repair criteria and to modify the SG tube 
inservice inspection frequency.
    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 continues to implement the Waterford 3 Steam 
Generator Program performance criteria for tube structural 
integrity, accident induced leakage, and operational leakage for the 
replacement SGs. Meeting the performance criteria provides 
reasonable assurance that the replacement SG tubing will remain 
capable of fulfilling its specific safety function of maintaining 
reactor coolant system (RCS) pressure boundary integrity throughout 
each operating cycle and in the unlikely event of a design basis 
accident.

[[Page 20634]]

    The Steam Generator Tube Rupture (SGTR) is the primary accident 
analysis associated with SG tube integrity. The replacement SG 
tubing contains improved materials that will reduce the likelihood 
of tubing flaws. The proposed change to remove alternate repair 
criteria from the SG inspection program does not affect the design 
of the replacement SGs, their method of operation, operational 
leakage limits, or primary coolant chemistry controls. Therefore, 
the proposed change does not affect the probability of a SGTR 
accident. The SGs will be designed with substantial margin to burst. 
The SG tube inspection repair limit will also identify potential 
flaws before they become a safety concern. The extension of the SG 
tube inspection frequency after initial inspection is based on the 
low likelihood of having potential tube flaws and is considered to 
be an acceptable inspection period to preserve pressure boundary 
integrity. As a result, there will be no affect on the previous dose 
analysis reported in the FSAR [Final Safety Analysis Report] and the 
consequences of any accident are unchanged.
    Therefore, this 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.
    Steam generator tube rupture events have already been postulated 
and analyzed in the Waterford 3 FSAR. The proposed change does not 
affect the design of the SGs, their method of operation, or primary 
or secondary coolant chemistry controls. Additionally, the proposed 
amendment does not impact any other plant systems or components. The 
TSs have established SG tube inspection requirements which assure 
that potential tubing flaws will be detected prior to affecting tube 
integrity and the RCS pressure boundary. Therefore, the proposed 
change does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    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 structural integrity, accident induced leakage, and 
operational leakage performance criteria required by the Waterford 3 
TSs provide substantial design margin for assuring SG tube integrity 
against the possibility of a SG tube pressure boundary failure. The 
proposed change removes an existing alternate repair criterion that 
is not applicable to the replacement SGs and establishes appropriate 
SG tube subsequent inspection periods consistent with the new SG 
tubing design. The replacement SGs will continue to meet their 
required performance criteria. The Waterford 3 SG tube inspection 
program will assure that this margin is maintained through the 
operational life of the plant.
    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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

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 15, 2010.
    Description of amendment request: This amendment request involves 
the adoption of Nuclear Regulatory Commission (NRC)-approved changes to 
the Standard Technical Specifications (STS) for Westinghouse plants 
(NUREG-1431), to allow relocation of specific TS surveillance 
frequencies to a licensee-controlled program. The proposed changes are 
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--Risk Informed Technical Specification Task Force (RITSTF) 
Initiative 5b,'' as announced in the Notice of Availability published 
in the Federal Register on July 6, 2009 (74 FR 31996). Additionally, 
the proposed changes would add a new program, the Surveillance 
Frequency Control Program, to TS Section 5, Administrative Controls. 
The changes are applicable to licensees using the probabilistic risk 
guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-
10, Revision 1, ``Risk-Informed Technical Specifications Initiative 5b, 
Risk-Informed Method for Control of Surveillance Frequencies.''
    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 adopted by the licensee is 
presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
Technical Specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    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 previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed 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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the Updated Final Safety Analysis Report and Bases 
to the Technical Specifications), because these are not affected by 
changes to the surveillance frequencies. Similarly, there is no 
impact to safety analysis acceptance criteria as described in the 
plant-licensing basis. To evaluate a change in the relocated 
surveillance frequency, EGC will perform a probabilistic risk 
evaluation using the guidance contained in NRC approved NEI 04-10, 
Revision 1 in accordance with the TS Surveillance Frequency Control 
Program. NEI 04-10, Revision 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the three standards of 10 
CFR 50.92(c) are

[[Page 20635]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

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

    Date of amendment request: February 15, 2010.
    Description of amendment request: This amendment request involves 
the adoption of Nuclear Regulatory Commission (NRC)-approved changes to 
the Standard Technical Specifications (STS) for Westinghouse plants 
(NUREG-1431), to allow relocation of specific TS surveillance 
frequencies to a licensee-controlled program. The proposed changes are 
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--Risk Informed Technical Specification Task Force (RITSTF) 
Initiative 5b,'' as announced in the Notice of Availability published 
in the Federal Register on July 6, 2009 (74 FR 31996). Additionally, 
the proposed changes would add a new program, the Surveillance 
Frequency Control Program, to TS Section 5, Administrative Controls. 
The changes are applicable to licensees using the probabilistic risk 
guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-
10, Revision 1, ``Risk-Informed Technical Specifications Initiative 5b, 
Risk-Informed Method for Control of Surveillance Frequencies.''
    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 adopted by the licensee is 
presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
Technical Specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    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 previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed 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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the Updated Final Safety Analysis Report and Bases 
to the Technical Specifications), because these are not affected by 
changes to the surveillance frequencies. Similarly, there is no 
impact to safety analysis acceptance criteria as described in the 
plant-licensing basis. To evaluate a change in the relocated 
surveillance frequency, EGC will perform a probabilistic risk 
evaluation using the guidance contained in NRC approved NEI 04-10, 
Revision 1 in accordance with the TS Surveillance Frequency Control 
Program. NEI 04-10, Revision 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
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. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

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

    Date of amendment request: February 4, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.3.61, ``Primary Containment 
Isolation Instrumentation,'' Table 3.3.6.1-1, ``Primary Containment 
Isolation Instrumentation,'' Function 6.a, ``Shutdown Cooling System 
Isolation, Recirculation Line Water Temperature--High,'' to enable 
implementation of a modification that replaces the temperature-based 
isolation instrumentation with reactor pressure-based isolation 
instrumentation. The proposed modification will address instrumentation 
reliability problems that have led to interruptions of Shutdown Cooling 
(SDC) system operation, leading to unplanned heat-up of reactor coolant 
while the reactor was in operational Modes 3 and 4.
    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 license amendment implements a revised process 
parameter and the associated Allowable Value (AV) for the DNPS Units 
2 and 3 SDC system isolation function 6.a in TS Table 3.3.6.1-1.
    The proposed changes to the isolation function do not affect the 
probability of any event initiators at the facilities. This 
isolation function is provided for equipment protection to prevent 
exceeding the system design temperature. The isolation function is 
not credited or assumed in the accident or transient analysis in the 
Updated Final Safety Analysis Report (UFSAR).
    The proposed changes will not degrade the performance of, or 
increase the number of challenges imposed on, safety-related 
equipment that is assumed to function during an accident situation. 
The SDC system and the isolation function that is being revised are 
not safety related and are not credited to function during an 
accident situation. The proposed changes will not alter any 
assumptions or change any mitigation actions in the radiological 
consequence evaluations in the UFSAR.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 20636]]

    (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 license amendment implements a revised process 
parameter and AV for the DNPS Units 2 and 3 SDC system isolation 
function 6.a in TS Table 3.3.6.1-1. The proposed change enables 
implementation of a modification that will enhance the reliability 
of instrumentation used to protect the functionality and integrity 
of the non safety-related SDC system. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints that initiate protective or mitigative actions. As a 
result, no new failure modes are being introduced.
    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 license amendment revises a process parameter and 
AV for the DNPS Units 2 and 3 SDC system isolation function 6.a in 
TS Table 3.3.6.1-1.
    The margin of safety is established through the design of the 
plant structures, systems, and components (SSCs), the parameters 
within which the plant is operated, and the setpoints for the 
actuation of equipment relied upon to respond to an accident.
    The proposed change to the SDC system isolation instrumentation 
function for the SDC system does not change the SSCs, operational 
parameters, or actuation setpoints for equipment that is relied upon 
to respond to an accident. Both the SDC system and the isolation 
function that is being revised are non-safety related and are not 
credited to function during an accident situation.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

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

    Date of amendment request: February 16, 2010.
    Description of amendment request: The proposed amendments would 
modify the DNPS Units 2 and 3, Technical Specifications (TS) by 
relocating specific surveillance frequencies to a licensee-controlled 
program with the adoption of Technical Specification Task Force (TSTF)-
425, ``Relocate Surveillance Frequencies to Licensee Control--Risk 
Informed Technical Specification Task Force (RITSTF) Initiative 5b,'' 
Revision 3. Additionally, the change would add a new program, the 
``Surveillance Frequency Control Program [SFCP],'' to TS Section 5, 
``Administrative Controls.''
    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. The licensee reviewed the proposed No Significant 
Hazards Consideration (NSHC) determination published in the Federal 
Register dated July 6, 2009 (74 FR 31996).
    The licensee has concluded that the proposed NSHC presented in the 
Federal Register notice is applicable to DNPS, Units 2 and 3. The 
proposed NSHC is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements (SRs) to licensee control under a 
new SFCP. Surveillance frequencies are not an initiator to any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased. The 
systems and components required by the TS for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the SRs, and be capable of performing 
any mitigation function assumed in the accident analysis. As a 
result, the consequences of any accident previously evaluated are 
not significantly increased.
    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 previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed 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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to the 
TS), because these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, EGC will 
utilize the guidance contained in NRC-approved NEI 04-10, in 
accordance with the TS SFCP. NEI 04-10, Revision 1 methodology 
provides reasonable acceptance guidelines and methods for evaluating 
the risk increase of proposed changes to surveillance frequencies 
consistent with Regulatory Guide 1.177.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

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

    Date of amendment request: February 15, 2010.
    Description of amendment request: The proposed amendments would 
modify the LaSalle County Station (LSCS) Technical Specifications (TS) 
by relocating specific surveillance frequencies to a licensee-
controlled program with the implementation of Nuclear Energy Institute 
(NEI) 04-10.
    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:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any

[[Page 20637]]

accident previously evaluated is not significantly increased. The 
systems and components required by the Technical Specifications for 
which the surveillance frequencies are relocated are still required 
to be operable, meet the acceptance criteria for the surveillance 
requirements, and be capable of performing any mitigation function 
assumed in the accident analysis. As a result, the consequences of 
any accident previously evaluated are not significantly increased.
    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 previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed 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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the Updated Final Safety Analysis Report and Bases 
to the Technical Specifications), because these are not affected by 
changes to the surveillance frequencies. Similarly, there is no 
impact to safety analysis acceptance criteria as described in the 
plant licensing basis. To evaluate a change in the relocated 
surveillance frequency, EGC will perform a probabilistic risk 
evaluation using the guidance contained in NRC approved NEI 04-10, 
Revision 1 in accordance with the TS Surveillance Frequency Control 
Program. NEI 04-10, Revision 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

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

    Date of amendment request: February 22, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.1.7, ``Standby Liquid Control (SLC) 
System,'' to extend the completion time associated with Condition B 
from 8 hours to 72 hours.
    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 amendment revises Technical Specification (TS) 
3.1.7, ``Standby Liquid Control (SLC) System,'' to extend the 
completion time (CT) associated with Condition B (i.e., ``Two SLC 
subsystems inoperable.'') from eight hours to 72 hours.
    The proposed change is based on a risk-informed evaluation 
performed in accordance with Regulatory Guides (RG) 1.174, ``An 
Approach for Using Probabilistic Risk Assessment in Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis,'' and RG 
1.177, ``An Approach for Plant-Specific, Risk-Informed Decision-
making: Technical Specifications.''
    The proposed amendment modifies an existing CT for a dual-train 
SLC system inoperability. The condition evaluated, the action 
requirements, and the associated CT do not impact any initiating 
conditions for any accident previously evaluated.
    The proposed amendment does not increase postulated frequencies 
or the analyzed consequences of an Anticipated Transient Without 
Scram (ATWS). Requirements associated with 10 CFR 50.62 will 
continue to be met. In addition, the proposed amendment does not 
increase postulated frequencies or the analyzed consequences of a 
large-break loss-of-coolant accident for which the SLC system will 
be used for pH control (i.e., upon NRC approval of an August 26, 
2008 proposed LSCS license amendment regarding the adoption of an 
alternate source term methodology). The extended CT provides 
additional time to implement actions in response to a dual-train SLC 
system inoperability, while also minimizing the risk associated with 
continued operation. 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 amendment revises TS 3.1.7 to extend the CT 
associated with Condition B from eight hours to 72 hours. The 
proposed amendment does not involve any change to plant equipment or 
system design functions. This proposed TS amendment does not change 
the design function of the SLC system and does not affect the 
system's ability to perform its design function. The SLC system 
provides a method to bring the reactor, at any time in a fuel cycle, 
from full power and minimum control rod inventory to a subcritical 
condition with the reactor in the most reactive xenon free state 
without taking credit for control rod movement. Required actions and 
surveillance requirements are sufficient to ensure that the SLC 
system functions are maintained. No new accident initiators are 
introduced by this amendment. Therefore, the proposed amendment 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 amendment revises TS 3.1.7 to extend the CT 
associated with Condition B from eight hours to 72 hours. The 
proposed amendment does not involve any change to plant equipment or 
system design functions. The margin of safety is established through 
the design of the plant structures, systems, and components, the 
parameters within which the plant is operated, and the setpoints for 
the actuation of equipment relied upon to respond to an event.
    Safety margins applicable to the SLC system include pump 
capacity, boron concentration, boron enrichment, and system response 
timing. The proposed amendment does not modify these safety margins 
or the point at which SLC is manually initiated, nor does it affect 
the system's ability to perform its design function. In addition, 
the proposed change complies with the intent of the defense-in-depth 
philosophy and the principle that sufficient safety margins are 
maintained, consistent with RG 1.177 requirements (i.e., Section C, 
``Regulatory Position,'' paragraph 2.2, ``Traditional Engineering 
Considerations'').

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

[[Page 20638]]

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

    Date of amendment request: February 16, 2010.
    Description of amendment request: The proposed amendments would 
modify the QCNPS Units 1 and 2, Technical Specifications (TS) by 
relocating specific surveillance frequencies to a licensee-controlled 
program with the adoption of Technical Specification Task Force (TSTF)-
425, ``Relocate Surveillance Frequencies to Licensee Control--Risk 
Informed Technical Specification Task Force (RITSTF) Initiative 5b,'' 
Revision 3. Additionally, the change would add a new program, the 
``Surveillance Frequency Control Program [SFCP],'' to TS Section 5, 
``Administrative Controls.''
    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. The licensee reviewed the proposed No Significant 
Hazards Consideration (NSHC) determination published in the Federal 
Register dated July 6, 2009 (74 FR 31996).
    The licensee has concluded that the proposed NSHC presented in the 
Federal Register notice is applicable to QCNPS, Units 1 and 2. The 
proposed NSHC is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements (SRs) to licensee control under a 
new SFCP. Surveillance frequencies are not an initiator to any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased. The 
systems and components required by the TS for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the SRs, and be capable of performing 
any mitigation function assumed in the accident analysis. As a 
result, the consequences of any accident previously evaluated are 
not significantly increased.
    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 previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed 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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to the 
TS), because these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, EGC will 
utilize the guidance contained in NRC-approved NEI 04-10, in 
accordance with the TS SFCP. NEI 04-10, Revision 1 methodology 
provides reasonable acceptance guidelines and methods for evaluating 
the risk increase of proposed changes to surveillance frequencies 
consistent with Regulatory Guide 1.177.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: December 14, 2009.
    Description of amendment request: The proposed amendment would 
remove the structural integrity requirements contained in Technical 
Specifications (TSs) 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) and their 
associated Bases; incorporate changes to accident monitoring 
instrumentation for consistency with NUREG-1432 actions and allowed 
outage times for conditions that drive a unit to hot shutdown; and 
administrative corrections based on obvious typos, previous amendments, 
or obsolete requirements.
    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 change to remove structural integrity controls 
from the TSs does not impact any mitigation equipment or the ability 
of the RCS [reactor coolant system] pressure boundary to fulfill any 
required safety function. The proposed change will continue to 
ensure the requirements of 10 CFR 50.55a are maintained as specified 
in TS 4.0.5 and the new administrative TS program for RCP [reactor 
coolant pump] flywheel inspections. The changes to the accident 
instrumentation actions and allowed outage time have no appreciable 
effect on accident initiation or mitigation. Since no other accident 
mitigation or initiators are impacted by this change, no design 
basis accidents are affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed change will not alter the plant configuration or 
change the manner in which the plant is operated. Structural 
integrity will continue to be maintained as required by 10 CFR 
50.55a and specified in TS 4.0.5 and the new administrative TS 
program for RCP flywheel inspections. Accident monitoring 
instrumentation does not contribute to failure modes. No new failure 
modes are being introduced by the proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Removing TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) from the 
TSs does not reduce the controls that are required to maintain the 
structural integrity of ASME Code Class 1, 2, or 3 components. There 
is no increase with any accident mitigation risk associated with the 
accident monitoring instrumentation TS changes as the proposed 
allowed outage times and the intervening step through HOT STANDBY 
are consistent with the equivalent to NUREG-1432 completion times 
and actions for post accident instrumentation and are equal to or 
more conservative than the current TS requirements. No other safety 
margins are impacted due to the proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

[[Page 20639]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & 
Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Acting Branch Chief: Douglas A. Broaddus.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: February 25, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement (SR) 
3.8.1.9, Diesel Generator (DG) Load Test, to correct a non-
conservative power factor (PF) value and to add a new note 
consistent with TS Task Force (TSTF) traveler TSTF-276-A, Revision 
2, ``Revise DG Full Load Rejection Test.''
    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.
    Performing a surveillance that tests the DG is not a precursor 
of any accident previously evaluated. Revising the PF limit to be 
more conservative, and relaxing the requirement to maintain PF when 
paralleled to offsite power does not significantly affect the method 
of performing the surveillances such that the probability of an 
accident would be affected. These changes only affect surveillances 
of mitigative equipment and, therefore, do not have an impact on the 
probability of an accident previously evaluated.
    Revising the surveillances by specifying a more conservative PF 
value ensures the DG's will provide the power assumed in 
calculations of design basis accident mitigation. Relaxing the 
requirement to maintain PF when paralleled to offsite power does not 
affect performance of the DG under accident conditions. The 
performance of the surveillances ensures that mitigative equipment 
is capable of performing its intended function, and therefore, the 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    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.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on a safety-related system or component and do not challenge 
the performance or integrity of safety related systems. As such, it 
does not introduce a mechanism for initiating a new or different 
accident than those described in the USAR [updated safety analysis 
report].
    Therefore, the proposed change does 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 proposed changes will continue to ensure the DGs are able to 
perform their design function as assumed in calculations that 
evaluate their function during design basis accidents. Decreasing 
the PF limit for testing will not affect the design or functioning 
of the DGs. The increased reactive loading required to maintain the 
PF below the limit is small and well within DG capability. Based on 
this, the ability of CNS [Cooper Nuclear Station] to mitigate the 
design basis accidents that rely on operation of the DG's is not 
adversely impacted. Revising the PF increases the margin of safety 
by specifying a more conservative value for the PF limit. Therefore, 
NPPD [Nebraska Public Power District] concludes these proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

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 System (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].

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: July 13, 2007, as supplemented 
by letters dated. July 13, 2007, September 30, 2008, March 5, 2009, 
March 23, 2009, March 1, 2010, and March 5, 2010.
    Brief description of amendment: The license amendment revises the 
Millstone Power Station, Unit No. 3 (MPS3) spent fuel pool (SFP) 
storage requirements. The July 13, 2007, license amendment request 
proposed a stretch power uprate (SPU) of MPS3. Included in a supplement 
dated July 13, 2007, was a request to amend the MPS3 SFP storage 
requirements. The July 13, 2007, request was noticed in the Federal 
Register on January 15, 2008 (73 FR 2549). By letter dated March 5, 
2008, Dominion Nuclear Connecticut, Inc. (DNC) separated the MPS3 SFP 
storage

[[Page 20640]]

requirements request from the MPS3 SPU request. The request to revise 
the MPS3 SFP storage requirements was re-noticed on September 8, 2009 
(74 FR 46241) using the original significant hazards consideration, 
specific to the request to revise the SFP storage.
    Date of issuance: March 26, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 248.
    Renewed Facility Operating License No. NPF-49: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: January 15, 2008 (73 FR 
2549) and September 8, 2009 (74 FR 46241). The supplemental letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination as 
published in the Federal Register (73 FR 2549). The SFP LAR no 
significant hazards consideration determination was noticed a second 
time, separate from the MPS3 SPU (74 FR 46241).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 2010.
    No significant hazards consideration comments received: No.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana 
Parish, Louisiana

    Date of amendment request: June 29, 2010.
    Brief description of amendment: The amendment revised the RBS 
Technical Specification (TS) 5.5.6, ``Inservice Testing Program.'' TS 
5.5.6 contains references to the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code, Section XI as the 
source for the inservice testing (IST) of ASME Code Class 1, 2, and 3 
pumps and valves. The proposed changes delete the references to Section 
XI of the ASME Code and incorporate references to the ASME Code for 
Operation and Maintenance of Nuclear Power Plants (OM Code). In 
addition, the amendment changes will limit applying Surveillance 
Requirement (SR) 3.0.2 to surveillances with a frequency of 2 years or 
less. These changes are consistent with the changes identified in the 
Improved Standard Technical Specifications (ISTS) in Technical 
Specification Task Force Traveler (TSTF) Change Travelers TSTF-479, 
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF-497, ``Limit 
Inservice Testing Program 3.0.2 Application to Frequencies of 2 Years 
or Less.''
    Date of issuance: March 31, 2010.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 25, 2009 (74 FR 
42928).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 2010.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station (GGNS), Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: October 27, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 2.1.1, ``Reactor Core SLs [Safety Limits],'' 
Subsection 2.1.1.2, to change the two recirculation loop safety limit 
for minimum critical power ratio (SLMCPR) from 1.08 to 1.09 and the 
single recirculation loop SLMCPR from 1.10 to 1.12. The changes to the 
TSs are necessary as a result of the GGNS Cycle 18 cycle-specific 
SLMCPR calculations.
    Date of issuance: March 25, 2010.
    Effective date: As of the date of issuance and shall be implemented 
after the current cycle (Cycle 17) is completed and prior to the 
operation of Cycle 18.
    Amendment No: 184.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 5, 2010 (75 FR 
461).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 25, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2 (Braidwood), Will County, Illinois, 
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2 
(Byron), Ogle County, Illinois

    Date of application for amendment: December 4, 2008, as 
supplemented by letters dated February 17, 2009; July 27, 2009; 
December 4, 2009; and January 29, 2010.
    Brief description of amendment: The amendments revise Technical 
Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, ``RCS [Reactor 
Coolant System] Specific Activity,'' and Surveillance Requirements 
3.4.16.1, 3.4.16.2, and 3.4.16.3. The revisions replace the current TS 
3.4.16 limit on RCS gross specific activity with a new limit on RCS 
noble gas-specific activity. The revisions adopt TS Task Force (TSTF) 
Change Traveler, TSTF-490, ``Deletion of E Bar Definition and Revision 
to RCS Specific Activity Tech Spec [sic],''
    Revision 0.
    Date of issuance: March 23, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: Braidwood Unit 1--162; Braidwood Unit 2--162; Byron 
Unit No. 1-167; and Byron Unit No. 2--167.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revise the TSs and Licenses.
    Date of initial notice in Federal Register: January 27, 2009 (74 FR 
4771).
    The 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 NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 23, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 7, 2009, as supplemented 
by letter dated October 5, 2009.
    Brief description of amendments: The amendments delete a footnote 
from DNPS Technical Specification (TS) 3.4.5, ``RCS Leakage Detection 
Instrumentation,'' that was incorporated as part of a limited duration 
emergency license amendment in August 2008, and is no longer 
applicable. The amendments also correct errors in the titles of 
analytical methods in DNPS and QCNPS TS 5.6.5, ``Core Operating Limits 
Report (COLR),'' paragraph b. The proposed changes delete historical 
analytical methods from DNPS and

[[Page 20641]]

QCNPS TS 5.6.5.b that are no longer applicable, and renumber the 
remaining analytical methods.
    Date of issuance: April 1, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 234/227, 246/241.
    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 30, 2009 (74 FR 
31322). The October 5, 2009, supplement, contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 1, 2010.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: September 28, 2009, as supplemented by 
letter dated January 20, 2010.
    Brief description of amendment request: The proposed amendment 
would support application of optimized weld overlays or full structural 
weld overlays. Applying these weld overlays on the reactor coolant pump 
suction and discharge nozzle dissimilar metal welds requires an update 
to the DBNPS leak-before-break (LBB) evaluation.
    Date of issuance: March 24, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 281.
    Facility Operating License No. NPF-3: The amendment revised the 
current licensing basis.
    Date of initial notice in Federal Register: February 22, 2010 (75 
FR 7628).
    The January 20, 2010 supplement, contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 24, 2010.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: November 6, 2008; superseded by 
letters dated August 4 and December 4, 2009.
    Brief description of amendment: The amendment modifies the Crystal 
River Unit 3 (CR-3) technical specifications (TS) surveillance 
requirements (SRs) related to allowable voltage and frequency limits 
for the emergency diesel generator (EDG) testing. Specifically, the 
amendment revises the CR-3 TS SRs 3.8.1.2, 3.8.1.6, 3.8.1.10.c.3 and 
3.8.1.10.c.4 to restrict the voltage and frequency limits for both slow 
and fast EDG starts.
    Date of issuance: December 10, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 236.
    Facility Operating License No. DPR-72: Amendment revises the 
facility operating license and the technical specifications.
    Date of initial notice in Federal Register: September 8, 2009 (74 
FR 46242). The supplement dated December 4, 2009, 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 December 10, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 4, 2009.
    Brief description of amendment: The amendment changed the Duane 
Arnold Energy Center Technical Specification (TS) Section 5.5.12 
(Primary Containment Leakage Rate Testing Program) to exclude the Main 
Steam pathway leakage contribution from the overall integrated leakage 
rate Type A test measurement and from the sum of the leakage rates from 
Type B and Type C tests and changed TS Section 3.6.1.3 (Primary 
Containment Isolation Valves) to remove the repair criterion for main 
steam isolation valves that fail their as-found leakage rate acceptance 
criterion found in current Surveillance Requirement 3.6.1.3.9.
    Date of issuance: March 31, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 276.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 2009 (74 FR 
31324).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 2010.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station (NMPNS), Unit No. 2 (NMP2), Oswego County, New 
York

    Date of application for amendment: June 29, 2009, as supplemented 
on August 13, 2009, and February 3, 2010.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.12, ``10 CFR 50 Appendix J Testing Program 
Plan,'' by replacing the reference to Regulatory Guide 1.163 with a 
reference to Nuclear Energy Institute (NEI) topical report NEI 94-01, 
Revision 2-A, as the implementation document used by NMPNS to develop 
the NMP2 performance-based leakage testing program in accordance with 
Option B of 10 CFR 50, Appendix J. In addition, the amendment allows 
NMPNS to extend the current interval for the NMP2 primary containment 
integrated leak rate test (ILRT) from 10 years to 15 years, and allows 
successive ILRTs to be performed at 15-year intervals.
    Date of issuance: March 30, 2010.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 134.
    Renewed Facility Operating License No. NPF-069: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: October 20, 2009 (74 FR 
53779).
    The supplemental letters dated August 13, 2009, and February 3, 
2010, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the Nuclear Regulatory Commission staff's initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 30, 2010.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station, 
Unit No. 1, Salem County, New Jersey

    Date of application for amendment: October 8, 2009, as supplemented 
by letter dated February 25, 2010.
    Brief description of amendments: The amendment approves a one-time 
change to Technical Specification (TS) 6.8.4.i,

[[Page 20642]]

``Steam Generator (SG) Program,'' regarding the SG tube inspection and 
repair required for the portion of the SG tubes passing through the 
tubesheet region. Specifically, for Salem Unit No. 1 refueling outage 
20 (planned for spring 2010) and subsequent operating cycles until the 
next scheduled SG tube inspection, the amendment limits the required 
inspection (and repair if degradation is found) to the portions of the 
SG tubes passing through the upper 13.1 inches of the approximate 21-
inch tubesheet region. In addition, the amendment revises TS 6.9.1.10, 
``Steam Generator Tube Inspection Report,'' to provide reporting 
requirements specific to the one-time change.
    Date of issuance: March 29, 2010.
    Effective date: As of the date of issuance, to be implemented prior 
to completion of refueling outage 20 (currently scheduled for spring 
2010).
    Amendment No.: 294.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendment 
revised the TSs and the License.
    Date of initial notice in Federal Register: January 5, 2010 (75 FR 
464).
    The letter dated February 25, 2010, provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2010.
    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 requests: February 3, 2009, and March 3, 2009; 
both applications were supplemented by letters dated November 20, 2009, 
and January 20, 2010.
    Brief description of amendments: The amendments approved a revision 
to the South Texas Project (STP), Units 1 and 2 Fire Protection Program 
for Fire Areas 27 and 31. In the event of a fire in the Fire Areas 27 
and 31, the amendments allow the licensee to perform operator manual 
actions to achieve and maintain safe shutdown in lieu of meeting the 
circuit separation and protection requirements of Title 10 of the Code 
of Federal Regulations, Part 50, Appendix R, Section III.G.2. The 
amendments revised the License Condition 2.E, ``Fire Protection,'' in 
the facility operating licenses, to reflect the changes. The approved 
changes to the Fire Protection Program will be documented in the 
licensee's ``Fire Hazards Analysis Report.''
    Date of issuance: March 31, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1--193; Unit 2--181.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses.
    Date of initial notices in Federal Register: August 25, 2009 (74 FR 
42929, 42930). The supplemental letters dated November 20, 2009, and 
January 20, 2010, provided additional information that clarified the 
applications, did not expand the scope of the applications 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 31, 2010.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to

[[Page 20643]]

Facility Operating License, 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 System's (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].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, any person(s) whose interest may be 
affected by this action may file a request for a hearing and a petition 
to intervene with respect to issuance of the amendment to the subject 
facility operating license. Requests for a hearing and a petition for 
leave to intervene shall be filed in accordance with the Commission's 
``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 
2. Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the Commission's PDR, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland, and electronically on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. 
If a request for a hearing or petition for leave to intervene is filed 
by the above date, 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 identify the specific contentions which the requestor/petitioner 
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 
requestor/petitioner 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
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.\1\ 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 to relief. A requestor/petitioner 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
requestor/petitioner seeks to adopt the contention of another 
sponsoring requestor/petitioner, the requestor/petitioner who seeks to 
adopt the contention must either agree that the sponsoring requestor/
petitioner shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring requestor/
petitioner a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
[email protected], or by telephone at (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 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or

[[Page 20644]]

representative, already holds an NRC-issued digital ID certificate). 
Based upon this information, the Secretary will establish an electronic 
docket for the hearing in this proceeding if the Secretary has not 
already established an electronic docket.
    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. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through EIE, users will be required to install a Web 
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser 
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant 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 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing 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 
E-Filing 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 using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
[email protected], or by a toll-free call at (866) 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, 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. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    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, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: March 22, 2010, as supplemented on March 
23, 2010.
    Description of amendment request: The previous Technical 
Specification (TS) 3.4.17, ``Chemical and Volume Control System 
(CVCS),'' Action B, allowed the licensee 24 hours to restore an 
inoperable makeup water pathway from the Refueling Water Storage Tank 
before taking further actions. This amendment increased the completion 
time of TS 3.4.17, Action B, from 24 hours to 72 hours for fuel cycle 
26.
    Date of issuance: March 25, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 223.
    Facility Operating License No. (DPR-23): Amendment revises the 
technical specifications.
    Public comments requested as to propose no significant hazards 
consideration (NSHC): No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, state consultation, and 
final NSHC determination are contained in a safety evaluation dated 
March 25, 2010.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Douglas A. Broaddus.

    Dated at Rockville, Maryland, this 12th day of April 2010.

    For The Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-8744 Filed 4-19-10; 8:45 am]
BILLING CODE 7590-01-P