Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 27825-27838 [2010-11564]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 95 (Tuesday, May 18, 2010)]
[Notices]
[Pages 27825-27838]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-11564]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0179]


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 that the Commission publish notice of any amendments issued, 
or proposed to be issued and grants the Commission the authority to 
issue and make immediately effective any amendment to an operating 
license upon a determination by the Commission that such amendment 
involves no significant hazards consideration, notwithstanding the 
pendency before the Commission of a request for a hearing from any 
person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 22 to May 5, 2010. The last biweekly 
notice was published on May 4, 2010 (75 FR 23808).

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) 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.

[[Page 27826]]

    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, 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 basis 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

[[Page 27827]]

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.
    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/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.
    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.

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

    Date of amendment request: March 29, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 5.5.7, ``Reactor Coolant Pump 
Flywheel Inspection Program,'' by extending the reactor coolant pump 
(RCP) motor flywheel inspection interval for certain RCP motors from 
the currently-approved 10-year inspection interval to an interval not 
to exceed 20 years. The availability of this TS revision was announced 
in the Federal Register on October 22, 2003 (68 FR 60422) as part of 
the consolidated line item improvement process. In its application, the 
licensee affirmed the applicability of the model no significant hazards 
consideration determination, as published in the Federal Register on 
June 24, 2003 (68 FR 37590).
    Basis for proposed no significant hazards consideration 
determination:

[[Page 27828]]

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:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change to the RCP flywheel examination frequency does 
not change the response of the plant to any accidents. The RCP will 
remain highly reliable and the proposed change will not result in a 
significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel during 
normal and accident conditions, the extremely low probability of a 
loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and 
assuming a conditional core damage probability (CCDP) of 1.0 (complete 
failure of safety systems), the core damage frequency (CDF) and change 
in risk would still not exceed the NRC's [Nuclear Regulatory 
Commission's] acceptance guidelines contained in RG 1.174 [Regulatory 
Guide 1.174, ``An Approach for Using Probabilistic Risk Assessment in 
Risk-Informed Decisions on Plant-Specific Changes to the Licensing 
Basis''] (<1.0E-6 per year). Moreover, considering the uncertainties 
involved in this evaluation, the risk associated with the postulated 
failure of an RCP motor flywheel is significantly low. Even if all four 
RCP motor flywheels are considered in the bounding plant configuration 
case, the risk is still acceptably low.
    The proposed change does not adversely affect accident initiators 
or precursors, nor alter the design assumptions, conditions, or 
configuration of the facility, or the manner in which the plant is 
operated and maintained; alter or prevent the ability of structures, 
systems, components (SSCs) from performing their intended function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits; or affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. Further, the proposed 
change does not increase the type or amount of radioactive effluent 
that may be released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposure. The proposed change 
is consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Accident Previously 
Evaluated
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does the 
change to examination frequency affect any existing accident scenarios, 
or create any new or different accident scenarios. Further, the change 
does not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or alter the methods 
governing normal plant operation. In addition, the change does not 
impose any new or different requirements or eliminate any existing 
requirements, and does not alter any assumptions made in the safety 
analysis. The proposed change is consistent with the safety analysis 
assumptions and current plant operating practice. Therefore, the 
proposed change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria are 
not impacted by this change. The proposed change will not result in 
plant operation in a configuration outside of the design basis. The 
calculated impact on risk is insignificant and meets the acceptance 
criteria contained in RG 1.174. There are no significant mechanisms for 
inservice degradation of the RCP flywheel. Therefore, the proposed 
change does 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 amendments involve 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. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: March 19, 2010.
    Description of amendment request: This submittal requests changes 
to extend the Technical Specification (TS) allowed outage time (AOT) 
for the Unit 1 and Unit 2 Suppression Pool Cooling (SPC) mode of the 
Residual Heat Removal (RHR) system, the Residual Heat Removal Service 
Water (RHRSW) system, the Emergency Service Water (ESW) system, and the 
A.C. Sources-Operating (Emergency Diesel Generators) from 72 hours to 
seven (7) days in order to allow for repairs of the RHRSW system 
piping. Specifically, the proposal adds a footnote to the affected TS 
limiting conditions for operation to indicate that the 72-hour AOT for 
the affected system may be extended once per calendar year, for one 
unit only, for a period of up to 7 days to allow for repairs of one 
RHRSW subsystem piping with the opposite unit shutdown, reactor vessel 
head removed and reactor cavity flooded, and other specific 
compensatory measures in effect.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee (Exelon) 
has provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS changes will not increase the probability of an 
accident since they will only extend the time period that one RHRSW 
subsystem, one loop of SPC, one ESW loop and two Emergency Diesel 
Generators (EDGs) can be out of service. The extension of the time 
duration that one RHRSW, one ESW loop and two EDGs are out of 
service has no direct physical impact on the plant. The proposed 
inoperable RHRSW subsystem, ESW loop and two EDGs are normally in a 
standby mode while the unit is in [Operational Condition] OPCON 1 or 
2 and are not directly supporting plant operation. Therefore, they 
can have no impact on the plant that would make an accident more 
likely to occur due to their inoperability.
    During transients or events which require these subsystems to be 
operating, there is sufficient capacity in the operable loops/
subsystems and available[,] but inoperable[,] equipment to support 
plant operation or shutdown. Therefore, failures that are accident 
initiators will not occur more frequently than previously postulated 
as a result of the proposed changes.

[[Page 27829]]

    In addition, the consequences of an accident previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR) will 
not be increased. With one RHRSW subsystem inoperable, one SPC loop, 
one ESW loop and two EDGs inoperable but verified available prior to 
entering the proposed configuration, a known quantity of equipment 
is inoperable. Based on the support functions of the RHRSW system, a 
review of the plant was performed to determine the impacts that the 
inoperable RHRSW subsystem would have on other systems. The impacts 
were identified for each system and it was determined whether there 
were any adverse effects on the systems. It was then determined how 
the adverse effects would impact each system's design basis and 
overall plant safety. The consequences of any postulated accidents 
occurring on Unit 1 or Unit 2 during these AOT extensions was found 
to be bounded by the previous analyses as described in the UFSAR. 
Since the inoperable ESW loop, selected emergency core cooling 
system (ECCS) pumps and EDGs will be verified available prior to 
entering the proposed configuration, they would have no impact on 
other systems.
    The minimum equipment required to mitigate the consequences of 
an accident and/or safely shut down the plant will be operable or 
available. Therefore, by extending certain AOTs and extending the 
assumptions concerning the combinations of events for the longer 
duration of each extended AOT, Exelon concludes that at least the 
minimum equipment required to mitigate the consequences of an 
accident and/or safely shut down the plant will still be operable or 
available during the extended AOT.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS changes will not create the possibility of a 
different type of accident since they will only extend the time 
period that one RHRSW subsystem and one loop of SPC can be out of 
service, and one ESW loop and two EDGs can be inoperable, but 
verified available, prior to entering the proposed configuration. 
The extension of the time duration that one RHRSW subsystem and one 
SPC loop is out of service, and one ESW loop and two EDGs are 
inoperable, but verified available, prior to entering the proposed 
configuration has no direct physical impact on the plant and does 
not create any new accident initiators. The systems involved are 
accident mitigation systems. All of the possible impacts that the 
inoperable equipment may have on its supported systems were 
previously analyzed in the UFSAR and are the basis for the present 
TS Action statements and AOTs. The impact of inoperable support 
systems for a given time duration was previously evaluated and any 
accident initiators created by the inoperable systems was evaluated. 
The lengthening of the time duration does not create any additional 
accident initiators for the plant.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The present RHRSW, SPC, ESW and EDG AOT limits were set to 
ensure that sufficient safety-related equipment is available for 
response to all accident conditions and that sufficient decay heat 
removal capability is available for a loss of coolant accident 
(LOCA) coincident with a loss of offsite power (LOOP) on one unit 
and simultaneous safe shutdown of the other unit. A slight reduction 
in the margin of safety is incurred during the proposed extended AOT 
due to the increased risk that an event could occur in a 7-day 
period versus a 72-hour period. This increased risk is judged to be 
minimal due to the low probability of an event occurring during the 
extended AOT and based on the following discussion of minimum ECCS/
decay heat removal requirements.
    The inoperable ESW loop, selected ECCS pumps and EDGs will be 
verified available prior to entering the proposed configuration; 
therefore, extension of the AOT will have no effect on the minimum 
ECCS equipment available or margin of safety.
    The reduction in the margin of safety from the extension of the 
RHRSW, SPC, ESW and EDG AOT limits is not significant since the 
remaining operable ECCS equipment is adequate to mitigate the 
consequences of any accident. This conclusion is based on the 
information contained in General Electric Company documents NEDO-
24708A, ``Additional Information Required for NRC Staff Generic 
Report on Boiling Water Reactors,'' Revision 1, dated December 1980, 
and NEDC[-]3093P-A, ``BWR Owner's Group Technical Specification 
Improvement Methodology (with Demonstration for BWR ECCS Activation 
Instrumentation),'' dated December 1988. These documents describe 
the minimum requirements to successfully terminate a transient or 
LOCA initiating event (with scram), assuming multiple failures with 
realistic conditions, and were used to justify certain TS AOTs per 
UFSAR Sections 6.3.1.1.2.o and 6.3.3.1. The minimum requirements for 
short-term response to an accident would be either one Low Pressure 
Coolant Injection (LPCI) pump or one Core Spray subsystem in 
conjunction with Automatic Depressurization System (ADS), or the 
High Pressure Coolant Injection (HPCI) system, which would be 
adequate to re-flood the vessel and maintain core cooling sufficient 
to preclude fuel damage. For long-term response, the minimum 
requirements would be one loop of RHR for decay heat removal, along 
with another low-pressure ECCS subsystem. These minimum requirements 
will be met since implementation of the proposed TS changes will 
require the operability or availability of HPCI, ADS, two LPCI 
subsystems (or one LPCI subsystem and one RHR subsystem during decay 
heat removal) and one Core Spray subsystem be maintained during the 
7-day period. Operations personnel are fully qualified by normal 
periodic training to respond to and mitigate a Design Basis 
Accident, including the actions needed to ensure decay heat removal 
while LGS Unit 1 and Unit 2 are in the operational configurations 
described within this submittal. Accordingly, procedures are already 
in place that address safe plant shutdown and decay heat removal for 
situations applicable to those in the proposed AOTs.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: March 24, 2010.
    Description of amendment request: The proposed amendment would 
modify the Three Mile Island, Unit 1 (TMI-1) Technical Specifications 
(TSs) by relocating specific surveillance frequencies to a new 
licensee-controlled program called the Surveillance Frequency Control 
Program. This change incorporates the adoption of Nuclear Energy 
Institute (NEI) 04-10, ``Risk-Informed Technical Specifications 
Initiative 5b, Risk-Informed Method for Control of Surveillance 
Frequencies,'' Revision (Rev.) 1. A description of the Surveillance 
Frequency Control Program will be added to the TMI-1 TSs.
    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 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

[[Page 27830]]

Program [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 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 [Nuclear Regulatory Commission] NRC) will continue to be met as 
described in the plant licensing basis (including the final safety 
analysis report and bases to TS), since 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, Exelon will perform a probabilistic risk 
evaluation using the guidance contained in NRC approved NEI 04-10, 
Rev. 1, in accordance with the TS SFCP. NEI 04-10, Rev. 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station (PBAPS), Unit 2, York and 
Lancaster Counties, Pennsylvania

    Date of amendment request: August 28, 2009, as supplemented by 
letter dated February 25, 2010.
    Description of amendment request: The proposed change would modify 
the PBAPS Unit 2 Technical Specification (TS) Section 5.5.12 to reflect 
a one-time extension of the Type A containment Integrated Leak Rate 
Test (ILRT) to no later than October 2015. The proposed TS revision 
would allow a one-time extension of 5 years to the 10-year frequency of 
the performance-based leakage rate testing program for the PBAPS Unit 2 
containment Type A ILRT 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. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves a one-time extension of the Primary 
Containment ILRT interval from 10 years to 15 years. The proposed 
change does not involve a physical change to the plant [* * *]. The 
Primary Containment function is to provide an essentially leak tight 
barrier against the uncontrolled release of radioactivity to the 
environment for postulated accidents. As such, the containment 
itself and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident, and do not involve any 
accident precursors or initiators. Therefore, the probability of 
occurrence of an accident previously evaluated is not significantly 
increased by the proposed change.
    Continued containment integrity is assured by the established 
programs for local leak rate testing and inservice/containment 
inspections, which are unaffected by the proposed change. As 
documented in NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' dated September 1995, industry experience has shown that 
local leak rate tests (Type B and C) have identified the vast 
majority of containment leakage paths, and that ILRTs detect only a 
small fraction of containment leakage pathways.
    The potential consequences of the proposed change have been 
quantified by analyzing the changes in risk that would result from 
extending the ILRT interval from 10 years to 15 years. Increasing 
the ILRT interval to 15 years for this one-time change is considered 
to be insignificant since it represents a very small change to the 
PBAPS, Unit 2 risk profile. Additionally, the proposed change 
maintains defense-in-depth by preserving a reasonable balance among 
prevention of core damage, prevention of containment failure, and 
consequence mitigation. PBAPS, Unit 2 has determined that the 
increase in conditional containment failure probability due to the 
proposed change is very small. Therefore, it is concluded that the 
proposed one-time extension of the Primary Containment ILRT interval 
from 10 years to 15 years does not significantly increase the 
consequences of an accident previously evaluated.
    Based on the above discussion, it is concluded that the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change involves a one-time extension of the Primary 
Containment ILRT interval. The containment and the testing 
requirements to periodically demonstrate the integrity of the 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors or initiators. The proposed change does not involve a 
physical change to the plant (i.e., no new or different type of 
equipment will be installed)[* * *].
    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?
    Response: No.
    The proposed one-time extension of the Primary Containment ILRT 
interval does not alter the manner in which safety limits, limiting 
safety system setpoints, or limiting conditions for operation are 
determined. The specific requirements and conditions of the 10 CFR 
50 Appendix J testing program plan, as defined in the Technical 
Specifications, exist to ensure that the degree of Primary 
Containment structural integrity and leak-tightness that is 
considered in the plant safety analyses is maintained. The overall 
containment leakage rate limit specified by the Technical 
Specifications is maintained, and Type B and C containment leakage 
tests will continue to be performed at the frequency currently 
required by the TS.
    Containment inspections performed in accordance with [the * * *] 
plant programs [described above] serve to provide a high degree of 
assurance that the containment will

[[Page 27831]]

not degrade in a manner that is detectable only by an ILRT. 
Furthermore, a risk assessment using the current PBAPS, Unit 2 
Probabilistic Risk Assessment internal events model concluded that 
extending the ILRT test interval from 10 years to 15 years results 
in a very small change to the PBAPS, Unit 2 risk profile.
    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 with the NRC staff changes noted in square brackets above, 
it appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: February 16, 2010.
    Description of amendment request: To revise the licensing bases by 
removing two technical specifications (TSs) that restrict movements of 
heavy loads over the spent fuel pools.
    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:

    TS 3/4.9.7, Crane Travel-Spent Fuel Storage Areas (reviewed for 
both units)
    FPL has evaluated whether or not a significant hazards 
consideration is involved with removing the TS 3/4.9.7, ``Crane 
Travel--Spent Fuel Storage Areas,'' from the Turkey Point Units 3 
and 4 TS by focusing on the three standards set forth in 10 CFR 
50.92, ``Issuance of amendment,'' as discussed below:
    (1) Would operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The removal of TS 3/4.9.7 will not increase the probability of a 
fuel handling accident (FHA), as evaluated in Chapter 14.2.1 of the 
UFSAR [Updated Final Safety Analysis Report], and is considered 
remote because of the administrative controls and physical 
limitations imposed on fuel handling operations. The load limit 
restriction, in conjunction with existing plant documents (for 
example, Turkey Point heavy load handling procedures) that restrict 
crane or other heavy load handling operations provide a defense-in-
depth approach to handling heavy loads in the spent fuel pool 
vicinity. The load limitation defined in TS 3/4.9.7 is preserved and 
will be implemented based on the operation limits and safety margins 
for the control of heavy loads consistent with NUREG-0612. The TS 
change does not represent any physical change to the plant systems, 
structures, or components. Therefore, the systems credited with 
mitigating the dose consequences of a FHA remain in place. The dose 
consequences of a fuel handling accident as discussed in Turkey 
Point UFSAR Chapter 14.2.1 will not increase because of the 
administrative controls and physical limitations imposed on fuel 
handling operations which minimize the likelihood of a FHA.
    Therefore, facility operation in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Would operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The removal of TS 3/4.9.7 does not represent any physical change 
to the plant systems, structures, or components. The same 
operational functions of moving new fuel, spent fuel, or other loads 
over the spent fuel pool are retained and therefore do not create or 
increase the possibility of a new or different kind of accident from 
any accident previously evaluated. Additionally, the load limit of 
2000 pounds over the spent fuel pool defined in TS 3/4.9.7 is 
preserved and implemented in existing plant documents and are 
established based on the operational limits and safety margins for 
the control of heavy loads consistent with NUREG-0612. Other 
measures which preclude the creation of a new or different type of 
accident include interlocks and physical stops, operator training, 
and load handling procedures.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Would operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The removal of TS 3/4.9.7 does not change the operational 
process of moving loads over the spent fuel pool. There are no 
changes to any physical plant systems, structures, or components. 
The spent fuel handling crane has weight sensors that are 
interlocked to limit the total load. In addition, an in-line weight 
sensing system is provided for each hoist to limit the lifting load 
to preclude accidental fuel damage should binding occur. When 
lifting over spent fuel, the total load is limited to 2000 pounds by 
current procedures, limit switches and load sensors. Because of 
these measures, no margin of safety is reduced or compromised.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

    Based on the above, FPL concludes that the proposed amendment does 
not involve a significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    TS 3/4.9.12, Handling of Spent Fuel Cask (reviewed for both units)
    FPL has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment of removing TS 3/
4.9.12, ``Handling of Spent Fuel Cask,'' by focusing on the three 
standards set forth in 10 CFR 50.92, ``Issuance of amendment,'' as 
discussed below:

    (1) Would operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The removal of TS 3/4.9.12 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The accident evaluated for the existing spent 
fuel cask handling crane is the drop of a single element cask as 
cited in UFSAR Section 14.2.1.3, ``Cask Drop Accident.'' This cask 
drop accident was analyzed and the radiological dose consequence, as 
a result of the cask drop, is determined to be within the limits of 
10 CFR 100. The current spent fuel cask handling crane at Turkey 
Point Units 3 and 4 has a single 105/15 ton main/auxiliary hook 
design capacity and is not designed as single-failure-proof. The new 
spent fuel cask handling crane will be single-failure-proof meeting 
all of the requirements of NUREG-0554, ``Single Failure Proof Cranes 
for Nuclear Power Plants'' and also NUREG-0612, Section 5.1.6, 
``Single Failure Proof Handling Systems.'' The probability of a cask 
drop accident using a single-failure-proof crane designed and 
operated to these NUREG requirements is considered to be extremely 
small.
    The design for the upgrade of the spent fuel cask handling crane 
is to increase the capacity to 130/25 tons (main/auxiliary hook). 
All crane components (hoist, bridge, girders, etc.) are designed and 
fabricated to retain control of and hold the maximum critical load 
(a planned 32 element spent fuel cask) in the unlikely event of the 
failure of a single component, coincident with a Design or Maximum 
earthquake.
    The objectives cited in Section 5.1 of NUREG-0612, ``Recommended 
Guidelines,'' for the control of heavy loads are satisfied. The 
probability of a cask drop accident using the new single-failure-
proof spent fuel cask crane, as compared to the existing non-single-
failure-proof crane, is therefore not increased. The increase of the 
consequences of an accident previously evaluated is also not 
increased because the potential for a cask drop by the new upgraded 
spent fuel cask handling crane is considered to be extremely small.

[[Page 27832]]

    Further, operational limits, interlocks, procedural and 
administrative controls, that restrict the handling of heavy loads 
over fuel stored in the spent fuel pool, provide additional defense-
in depth to ensure that a load could not be dropped that would 
result in dose consequences greater than previously evaluated.
    It is concluded that facility operation in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Would operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    Operation of the spent fuel cask handling crane after the 
upgrade to a single-failure-proof design will remain the same as the 
operation of the existing spent fuel cask handling crane. The 
distinction is the load that will be lifted.
    The new spent fuel cask is a multiple assembly cask, in contrast 
to a single assembly cask as currently specified for use. The 
current spent fuel cask handling crane is designed to lift a single 
element spent fuel cask. The upgraded capacity of the new spent fuel 
cask handling crane will allow for lifting a cask designed to hold a 
maximum of 32 spent fuel assemblies. Current operating and 
administrative procedures that restrict the movement of heavy loads 
over fuel stored in the spent fuel pool remain in place. The new 
spent fuel cask handling crane is designed, fabricated and tested to 
single-failure-proof requirements (NUREG-0554, ``Single Failure 
Proof Cranes for Nuclear Power Plants'' and NUREG-0612, Section 
5.1.6, ``Single Failure Proof Handling Systems'') and will be 
operated within the procedural and administrative framework as the 
currently installed spent fuel cask handling crane. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created from the removal of TS 3/4.9.12.
    Therefore, it can be concluded that the operation of the 
facility in accordance with the proposed amendment would not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (3) Would operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The existing spent fuel cask handling crane is not designed as 
single-failure-proof in accordance with NUREG-0612. The new spent 
fuel cask handling crane is designed, and will be fabricated, 
installed and tested to the single-failure-proof requirements as 
outlined in NUREG-0612, Section 5.1.6, ``Single Failure Proof 
Handling Systems.'' The use of the defense-in-depth approach for the 
control and handling of heavy loads as cited in Section 5.1 of 
NUREG-0612, ``Recommended Guidelines,'' provides assurance that 
there is a sufficient margin of safety in the handling of heavy 
loads. Thereby, the removal of TS 3/4.9.12 will not involve a 
significant reduction in the margin of safety.
    Defense-in-depth measures include operational limits, 
interlocks, procedural and administrative controls, rigging, load 
paths, testing, training, maintenance and other related 
considerations. These measures provide assurance that the margin of 
safety is not reduced in the operation of the facility by meeting 
all the requirements of NUREG-0612 and NUREG-0554. The specific 
requirements and FPL compliance with them is documented in the 
NUREG-0554 Compliance Matrix [Attachment 3 to this application].
    The design for the upgrade of the spent fuel cask handling crane 
is to increase the capacity to 130/25 tons (main/auxiliary hook). 
The spent fuel cask handling crane has a Main Hoist and Auxiliary 
Hoist Cable Safety Factor of a minimum 10:1 on nominal breaking 
strength at 130 tons and 25 tons respectively and is fully compliant 
with ASME NOG-1 Section 5425.1. The Main Hoist Hook and Auxiliary 
Hoist Hook Safety Factor have a 10:1 minimum on ultimate strength at 
130 tons and 25 tons, respectively.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

    Based on the above, FPL concludes that the proposed amendment does 
not involve a significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

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

    Date of amendment request: January 14, 2010.
    Description of amendment request: The amendments would revise a 
number of Technical Specification (TS) requirements, to impose similar 
restrictions on the movement of non-irradiated fuel assemblies to those 
currently in place for movement of irradiated fuel assemblies. The 
additional restrictions will limit the movement of all fuel assemblies 
over irradiated fuel assemblies in containment or in the fuel storage 
pool. The affected TS Limiting Conditions for Operation (LCOs) are: LCO 
3.3.8, ``Containment Purge Isolation Signal (CPIS),'' LCO 3.3.9, 
``Control Room Isolation Signal (CRIS),'' LCO 3.7.11, ``Control Room 
Emergency Air Cleanup System (CREACUS),'' LCO 3.7.16, ``Fuel Storage 
Pool Water Level,'' LCO 3.8.2, ``AC Sources--Shutdown,'' LCO 3.8.5, 
``DC Sources--Shutdown,'' LCO 3.8.8, ``Inverters--Shutdown,'' LCO 
3.8.10, ``Distribution Systems--Shutdown,'' LCO 3.9.3, ``Containment 
Penetrations,'' and LCO 3.9.6, ``Refueling Water Level.''
    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.
    This proposed change revises Technical Specifications 
applicability wording regarding the movement of fuel assemblies in 
containment and the fuel storage pool at the San Onofre Nuclear 
Generating Station (SONGS) Units 2 and 3 to include the movement of 
both irradiated and non-irradiated fuel assemblies. The proposed 
applicability is more comprehensive than the current Applicability.
    Expanding the applicability of the relevant Technical 
Specifications is necessary to account for updated fuel drop 
analyses which demonstrate that impacted spent fuel assemblies may 
be damaged. Consequently, movement of non-irradiated fuel assemblies 
could result in a Fuel Handling Accident that has radiological 
consequences. Changing the applicability of the relevant Technical 
Specifications does not affect the probability of a Fuel Handling 
Accident. The expanded applicability provides assurance that 
equipment designed to mitigate a Fuel Handling Accident is capable 
of performing its specified safety function, such that the 
consequences of an accident are not increased.
    Consequently, 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 revised spent fuel drop analyses demonstrate that impacted 
fuel assemblies may be damaged as the result of a dropped fuel 
assembly. The existing SONGS Technical Specifications regarding 
movement of fuel assemblies are not applicable for movement of non-
irradiated fuel assemblies. A drop of a non-irradiated fuel assembly 
that has radiological consequences could occur during periods when 
equipment that would be required to mitigate those consequences is 
not required

[[Page 27833]]

to be OPERABLE in accordance with the existing Technical 
Specifications.
    The proposed changes to the Technical Specifications 
applicability language regarding the movement of fuel assemblies in 
containment and the fuel storage pool at SONGS Units 2 and 3 ensure 
that Limiting Conditions of Operation and appropriate Required 
Actions for required equipment are in effect during fuel movement. 
This provides assurance that any Fuel Handling Accident that may 
occur will remain within the initial assumptions of accident 
analyses.
    Consequently, there is no possibility of a new or different kind 
of accident due to this change.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed Technical Specifications change will not affect 
protection criterion for plant equipment and will not reduce the 
margin of safety. By extending the Applicability to the movement of 
non-irradiated fuel assemblies, the current margin of safety is 
maintained.
    Consequently, there is no significant reduction in a margin of 
safety due to this change.

    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: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Michael T. Markley.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: November 25, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation,'' that would add a new 
Required Action Q.1 to require restoration of an inoperable Balance of 
Plant (BOP) ESFAS train to OPERABLE status within 24 hours. In 
addition, the Completion Times for TS 3.3.2 Required Actions J.1 and 
O.1 to trip inoperable channels that provide inputs to BOP ESFAS would 
also be extended to 24 hours. Shutdown track Completion Times to be in 
MODES 3 and 4 would be increased to reflect longer restoration times. 
Separate Condition entry for TS Condition J would be restricted to 
assure that Function 6.g in TS Table 3.3.2-1 will provide a start 
signal to the motor-driven auxiliary feedwater pumps from one train of 
BOP ESFAS actuation logic.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed to the protection systems. The same 
reactor trip system (RTS) and engineered safety feature actuation 
system (ESFAS) instrumentation will continue to be used. The 
protection systems will continue to function in a manner consistent 
with the plant design basis. There will be no changes to the BOP 
ESFAS surveillance and operating limits.
    The proposed changes will not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes will not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended functions to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits.
    The proposed changes do not affect the way in which safety-
related systems perform their functions.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR [Final Safety Analysis Report].
    The applicable radiological dose acceptance criteria will 
continue to be met.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no proposed changes in the method by which any safety-
related plant SSC performs its safety function. The proposed changes 
will not affect the normal method of plant operation or change any 
operating parameters. No equipment performance requirements will be 
affected. The proposed changes will not alter any assumptions made 
in the safety analyses.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be 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.
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, Solid State Protection S