Biweekly Notice: Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 20627-20644 [2010-8744]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 75 (Tuesday, April 20, 2010)]
[Notices]
[Pages 20627-20644]
From the Federal Register Online via the Government Printing 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, https://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 
hearing.docket@nrc.gov, 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 https://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 
https://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 https://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 https://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  https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, 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 
https://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, https://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 pdr.resource@nrc.gov.

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