[Federal Register Volume 75, Number 95 (Tuesday, May 18, 2010)]
[Notices]
[Pages 27825-27838]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-11564]
=======================================================================
-----------------------------------------------------------------------
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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the 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
[email protected], or by telephone at (301) 415-1677, to request
(1) a digital ID certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or
[[Page 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 http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home addresses, or home phone numbers in their
filings, unless an NRC regulation or other law requires submission of
such information. With respect to copyrighted works, except for limited
excerpts that serve the purpose of the adjudicatory filings and would
constitute a Fair Use application, participants are requested not to
include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS should contact the NRC PDR Reference staff at 1 (800)
397-4209, (301) 415-4737, or by e-mail to [email protected].
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 System, BOP ESFAS, MSFIS [main steam/
feedwater isolation system], or LSELS [load shedder and emergency
load sequencer] used in the plant protection systems.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F[Delta]H), loss of coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other margin of safety. The applicable radiological
dose consequence acceptance criteria will continue to be met.
The proposed changes do not eliminate any surveillances or alter
the frequency of surveillances required by the Technical
Specifications. No instrument setpoints or system response times are
affected. None of the acceptance criteria for any accident analysis
will be changed.
The proposed changes will have no impact on the radiological
consequences of a design basis accident.
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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 30, 2010.
Description of amendment request: The proposed amendments would
modify the North Anna Technical Specifications (TSs) by relocating
specific surveillance frequencies to a licensee-controlled program with
the implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk-
Informed Technical Specifications Initiative 5b, Risk-Informed Method
for Control of Surveillance Frequencies.'' The changes
[[Page 27834]]
are consistent with NRC-approved Industry Technical Specifications Task
Force (TSTF) Standard Technical Specifications (STS) change TSTF-425,
Revision 3. The Federal Register notice published on July 6, 2009 (74
FR 31996), announced the availability of this TS improvement.
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 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 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, Dominion
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Gloria Kulesa.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices, either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of application for amendments: June 25, 2008, as supplemented
on November 6, 2008, March 9, 2009, June 12, 2009, December 18, 2009,
and March 26, 2010.
Brief description of amendment request: The proposed amendment
would revise the PBAPS, Units 2 and 3, Technical Specification Section
4.3.1.1.a concerning the spent fuel pool k-infinity value.
Date of publication of individual notice in Federal Register: April
26, 2010 (75 FR 21680).
Expiration date of individual notice: May 26, 2010 (comment
request); June 25, 2010 (hearing request).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action, see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management System (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are
[[Page 27835]]
problems in accessing the documents located in ADAMS, contact the PDR
Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to
[email protected].
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: November 23, 2009, as
supplemented by letter dated April 26, 2010.
Brief description of amendment: The license amendment request
revises the Millstone Power Station, Unit 3 (MPS3) Technical
Specification (TS) 6.8.4.g, ``Steam Generator Program,'' to exclude a
portion of the tubes below the top of the steam generator tubesheet
from periodic steam generator tube inspections. This request also
removes reference to the previous Cycle 13 interim alternate repair
criteria.
Date of issuance: May 3, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 249.
Renewed Facility Operating License No. NPF-49: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: January 26, 2010 (75 FR
4114). The supplemented dated April 26, 2010, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 3, 2010.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: December 1, 2008.
Brief description of amendments: The amendments correct a non-
conservative Technical Specification (TS) Surveillance Requirement by
revising McGuire TS 3.8.1.4 to increase the minimum required amount of
fuel oil for the Emergency Diesel Generators fuel oil day tank as read
on the local fuel gauge used to perform the surveillance.
Date of issuance: May 5, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 254 and 234.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: May 19, 2009 (74 FR
23442).
The supplements dated July 30, 2009, December 2, 2009, and March
10, 2010, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 5, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of application for amendments: August 7, 2008, as supplemented
on May 7, 2009, and January 19, 2010.
Brief description of amendments: The August 7, 2008, submittal
contained several areas of review that are being dispositioned as
separate amendment requests. The amendments associated with this notice
revise the PBAPS Units 2 and 3 Technical Specifications (TS) to delete
the list of emergency diesel generator critical trips from TS
Surveillance Requirement (SR) 3.8.1.13 and clarify that the purpose of
the SR is to verify that the non-critical trips are bypassed. This TS
change adopts Technical Specification Task Force (TSTF) Traveler 400,
Revision 1, ``Clarify SR on Bypass of DG [diesel generator] Automatic
Trips.''
Date of issuance: April 30, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 275 and 279.
Renewed Facility Operating License Nos. DPR-44 and DPR-56:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: May 5, 2009 (74 FR
20744).
The supplements dated May 7, 2009, and January 19, 2010, clarified
the application, did not expand the scope of the application as
originally noticed, and did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 30, 2010.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: April 2, 2009.
Brief description of amendments: The amendment revised Technical
Specification (TS) 3.3.1 entitled, ``Reactor Trip System (RTS)
Instrumentation'' to add Surveillance Requirement 3.3.1.16 to Function
3 of TS Table 3.3.1-1 to verify that the RTS response times are within
limits every 18 months on staggered basis. The change is based on a
reanalysis of the Rod Cluster Control Assembly Bank Withdrawal at Power
event.
Date of issuance: April 26, 2010.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1--151; Unit 2-151.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 19, 2009 (74 FR
23446).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 26, 2010.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date
[[Page 27836]]
the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to
[email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, any person(s) whose interest may be
affected by this action may file a request for a hearing and a petition
to intervene with respect to issuance of the amendment to the subject
facility operating license. Requests for a hearing and a petition for
leave to intervene shall be filed in accordance with the Commission's
``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part
2. Interested person(s) should consult a current copy of 10 CFR 2.309,
which is available at the Commission's PDR, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland, and electronically on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A requestor/petitioner
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
[[Page 27837]]
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
requestor/petitioner seeks to adopt the contention of another
sponsoring requestor/petitioner, the requestor/petitioner who seeks to
adopt the contention must either agree that the sponsoring requestor/
petitioner shall act as the representative with respect to that
contention, or jointly designate with the sponsoring requestor/
petitioner a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the Internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the participant should
contact the Office of the Secretary by e-mail at
[email protected], or by telephone at (301) 415-1677, to request
(1) a digital ID certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant
[[Page 27838]]
or party to use E-Filing if the presiding officer subsequently
determines that the reason for granting the exemption from use of E-
Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
social security numbers, home addresses, or home phone numbers in their
filings, unless an NRC regulation or other law requires submission of
such information. With respect to copyrighted works, except for limited
excerpts that serve the purpose of the adjudicatory filings and would
constitute a Fair Use application, participants are requested not to
include copyrighted materials in their submission.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: March 29, 2010, as supplemented
by letters dated March 29 and April 26, 2010.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' Condition J under function 6.g in TS Table
3.3.2-1. Function 6.g provides an auxiliary feedwater (AFW) start
signal that is provided to the motor-driven AFW pumps in the event of a
trip of both turbine-driven main feedwater (MFW) pumps. The licensee
determined that the design and normal operation of the MFW pumps could
result in a condition that does not conform to TS Table 3.3.2-1,
function 6.g. Entry into Limiting Condition for Operation (LCO) 3.0.3
will be required; therefore, the TS change was needed to address this
condition. The change to Condition J allows placing the two channels in
a tripped condition on one MFW pump when placing the pump into service
or removing the pump from service prior to resetting the MFW pump. With
the revision to Condition J, the licensee will not require an entry
into LCO 3.0.3. Specifically, the changes revised Condition J for ESFAS
instrumentation function 6.g to read, ``One or more Main Feedwater
Pumps trip channel(s) inoperable,'' made corresponding changes to
Required Action J.1, and placed a Note above Required Actions J.1 and
J.2 for consistency with the revised Condition.
Date of issuance: May 5, 2010.
Effective date: As of its date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 196.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (75 FR 19431; April 14, 2010).
The supplemental letters dated March 29 and April 26, 2010,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed NSHC determination as
published in the Federal Register. The notice provided an opportunity
to submit comments on the Commission's proposed NSHC determination. No
comments have been received. The notice also provided an opportunity to
request a hearing by June 14, 2010, but indicated that if the
Commission makes a final NSHC determination, any such hearing would
take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated May 5, 2010.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Dated at Rockville, Maryland, this 6th day of May 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-11564 Filed 5-17-10; 8:45 am]
BILLING CODE 7590-01-P