[Federal Register Volume 74, Number 65 (Tuesday, April 7, 2009)]
[Notices]
[Pages 15765-15778]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-7494]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0148]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 12, 2009 to March 25, 2009. The last
biweekly notice was published on March 24, 2009 (74 FR 12390).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch, 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. Copies of written comments
received may be examined at the Commission's Public Document Room
(PDR), located at One White Flint North, Public File Area 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 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/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
[[Page 15766]]
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
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, which the NRC
promulgated in August 28, 2007 (72 FR 49139). 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 a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer \TM\ to
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer \TM\ is free and is available
at http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
Information about applying for a digital ID certificate is available on
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday, excluding government holidays. The help
electronic filing Help Desk can be contacted by telephone at 1-866-672-
7640 or by e-mail at [email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) First class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained
[[Page 15767]]
absent a determination by the Commission, the presiding officer, or the
Atomic Safety and Licensing Board that the petition and/or request
should be granted and/or the contentions should be admitted, based on a
balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).
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, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings, 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.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, 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].
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 2, 2008.
Description of amendment request: The amendments would revise the
technical specifications to allow manual operation of the containment
spray system and to revise the upper and lower limits of the refueling
water storage tank.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (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 Containment Spray System and RWST [refueling water storage
tank] are accident mitigation equipment. As such, changes in operation
of these systems cannot have an impact on the probability of an
accident.
The RWST will continue to comply with all applicable regulatory
requirements and design criteria following approval of the proposed
changes (e.g., train separation, redundancy, and single failure). The
water level on the containment floor will be higher at the start of
transfer to the containment sump but will remain below the maximum
design level analyzed for equipment submergence. The change in the sump
pH will not result in a significant increase in radiological
consequences of a LOCA [loss of coolant accident]. Therefore, the
design functions performed by the equipment are not changed.
The delay in containment spray operation will result in an increase
in containment temperature, containment pressure, offsite dose, and
control room dose during a LOCA or high energy line break inside
containment. Containment analyses have been performed to demonstrate
that containment pressure and temperature remain within the design
limits and there is no significant impact on the environmental
qualification for equipment inside containment. The impact on piping
and supports is acceptable without modification. The reduction in
fission product removal due to delayed containment spray operation does
not result in exceeding the offsite dose and control room dose limits
in 10 CFR 50.67 and 10 CFR Part 50, Appendix A, GDC 19. The analysis of
the change in containment conditions due to a single failure of an
operating spray pump and the suspension of containment spray determined
that the pressure remained below the design limits.
Regarding the proposed change to adopt TSTF-493, Rev. 3 on a
limited basis, the change clarifies the requirements for
instrumentation to ensure the instrumentation will actuate as assumed
in the safety analysis. Instruments are not an assumed initiator of any
accident previously evaluated. As a result, the proposed change will
not increase the probability of an accident previously evaluated. The
proposed change will ensure that the instruments actuate as assumed to
mitigate the accidents previously evaluated. As a result, the proposed
change will not increase the consequences of an accident previously
evaluated.
Based on this discussion, the proposed amendment does not
significantly increase 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 modifications to install RWST narrow range level indication
will be seismically qualified and isolated from the safety related
portion of the RWST level indication system. As such, the new level
indication will not create the possibility of a new or different kind
of accident.
The modification to the low level setpoint will not install any new
plant equipment. The setpoint will continue to be included within the
engineered safeguards features instrumentation and monitored according
to the applicable surveillance requirements. The evaluation of the new
level setpoint and the change in the swapover sequence concluded that
the equipment aligned to the sump will continue to have sufficient
suction pressure prior to containment sump suction swapover. The design
of the RWST low level instrumentation-complies with all applicable
regulatory requirements and design criteria.
The overall function of the Containment Spray System is not changed
by this proposed amendment. The proposed change alters the method of
controlling the safety system following a design basis event so that
manual actions are substituted for automatic actions. Calculations
confirm that these actions will be taken within the appropriate
scenario sequence timing to provide containment cooling and source term
reduction with no significant increase in radiological consequences and
without exceeding containment design limits.
Regarding the proposed change to adopt TSTF-493, Rev. 3 on a
limited basis, the change does 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. The change
does not alter assumptions made in the safety analysis but ensures that
the instruments behave as assumed in the accident analysis. The
proposed change is consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility of a
new or
[[Page 15768]]
different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
The proposed change has the potential to increase the radiological
dose at the site boundary and in the control room. However, the
calculations demonstrate that the dose consequences at the site
boundary, low population zone, and control room remain within
regulatory acceptance limits. Additional analysis concluded:
Peak containment pressure for analyzed design basis
accidents will not be significantly increased and containment design
limits will not be exceeded.
Assumptions used in the environmental qualification of
equipment exposed to the containment atmosphere remain bounding.
Pumps aligned to the RWST and to the containment sump will
have adequate suction pressure.
Regarding the proposed change to adopt TSTF-493, Rev. 3 on a
limited basis, the change clarifies the requirements for
instrumentation to ensure the instrumentation will actuate as assumed
in the accident analysis. No change is made to the accident analysis
assumptions and no margin of safety is reduced as part of this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: October 2, 2008.
Description of amendment request: The amendments would revise
Technical Specifications (TS) associated with the verification of ice
condenser door operability. The proposed amendment affects the current
TS surveillance requirements 3.6.13.5 and 3.6.13.6.
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 only analyzed accidents of possible consideration in regards to
changes potentially affecting the ice condenser are a loss of coolant
accident (LOCA) and a high energy line break (HELB) inside Containment.
However, the ice condenser is not postulated as being the initiator of
any LOCA or HELB. This is because it is designed to remain functional
following a design basis earthquake, and the ice condenser does not
interconnect or interact with any systems that interconnect or interact
with the Reactor Coolant or Main Steam Systems. Since these proposed
changes do not result in, or require, any physical change to the ice
condenser that could introduce an interaction with the Reactor Coolant
or Main Steam Systems, then there can be no change in the probability
of an accident previously evaluated. Regarding consequences of analyzed
accidents, the ice condenser is an engineered safety feature designed,
in part, to limit the Containment sub-compartment and Containment
vessel pressure immediately following the initiation of a LOCA or HELB.
Conservative sub-compartment and Containment pressure analysis shows
these criteria will be met if the total ice mass within the ice bed is
maintained in accordance with the DBA analysis; therefore, the proposed
TS [Technical Specification] SR [surveillance requirement] changes of
these requirements will not increase the consequences of any accident
previously evaluated.
Thus, based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
As previously described, the ice condenser is not postulated as
being the initiator of any design basis accident. The proposed changes
do not impact any plant system, structure or component that is an
accident initiator. The proposed TSs and TS Bases changes do not
involve any hardware changes to the ice condenser or other change that
could create any new accident mechanisms. Therefore, there can be no
new or different accidents created from those already identified and
evaluated
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of the
fission product barriers to perform their design functions during and
following an accident situation. These barriers include the fuel
cladding, the reactor coolant system, and the Containment system. The
performance of the fuel cladding and the reactor coolant system will
not be impacted by the proposed changes. The Application provides a
description of additional sub-compartment and Containment pressure
response analysis that has been performed. This analysis demonstrates
that Containment will remain fully capable of performing its design
function with implementation of the proposed changes. Therefore, no
safety margin will be significantly impacted.
The changes proposed in this LAR [license amendment request] do not
make any physical alteration to the ice condenser doors, nor does it
affect the required functional capability of the doors in any way. The
intent of the proposed changes to the ice condenser door surveillance
requirements is to eliminate an unnecessary and overly restrictive
Lower Inlet Door torque surveillance test. There will be no degradation
in the operable status of the ice condenser doors and the ability to
confirm operability for the ice condenser doors will be maintained,
such that the doors will continue to fully perform their safety
function as assumed in the plant's safety analyses.
Thus, it can be concluded that the proposed TS and TS Bases changes
do not involve a significant reduction in the 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
[[Page 15769]]
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: October 8, 2008.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) by removing and updating portions of the
TSs which are out of date or are obsolete including footnotes and
references.
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 changes are administrative in nature and therefore
they do not involve any change in the design, configuration, or
operation of the nuclear units. All Limiting Conditions for Operation,
Limiting Safety System Settings and Safety Limits specified in the
Technical Specifications remain unchanged. The Physical Security and
related plans, Operator Training and Requalification Programs, Quality
Assurance Programs, and the Emergency Plans will not be materially
changed by the proposed license amendment due to its administrative
nature.
The technical qualifications of the operating licensee will not be
reduced. Personnel engaged in operation, maintenance, engineering,
assessment, training, and other related services will not be changed.
Duke officers and executives currently responsible for the overall safe
operation of the nuclear plants are expected to continue in the same
capacity.
Therefore, the proposed amendment does not involve an increase in
the probability or consequences of an accident previously analyzed.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative in nature and therefore
they do not involve any change in the design, configuration, or
operation of the nuclear plant. The current plant safety analyses,
therefore, remain complete and accurate in addressing the design basis
events and in analyzing plant response and consequences.
The Limiting Conditions for Operations, Limiting Safety System
Settings and Safety Limits specified in the Technical Specifications
are not affected by the proposed changes. As such, the plant conditions
for which the design basis accident analyses were performed remain
valid.
The amendment does not introduce a new mode of plant operation or
new accident precursors, does not involve any physical alterations to
plant configurations or make changes to system set points that could
initiate a new or different kind of accident.
Therefore, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are administrative in nature and therefore
they do not involve a change in the design, configuration, or operation
of the nuclear plants. The change does not affect either the way in
which the plant, structures, systems, and components perform their
safety function or their design and licensing bases.
Plant safety margins are established through Limiting Conditions
for Operation, Limiting Safety System Settings and Safety Limits
specified in the Technical Specifications. Because there is no change
to the physical design of the plant, there is no change to any of these
margins.
Therefore, the proposed amendment does not involve a significant
reduction in the 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: October 14, 2008.
Description of amendment request: The amendments would revise the
Technical Specification [TS] Administrative Controls, ``Inservice
Testing Program,'' for consistency with the requirements of Title 10 of
the Code of Federal Regulations (10 CFR) 50.55a(f)(4) for pumps and
valves which are classified as American Society of Mechanical Engineers
[ASME] Code Class 1, Class 2, and Class 3.
Basis for proposed no significant hazards consideration
determination: As required by
10 CFR 50.91(a), the licensee has provided its analysis of the
issue of no significant hazards consideration, which is presented
below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4)
regarding the inservice testing of pumps and valves which are
classified as ASME Code Class 1, Class 2, and Class 3. The proposed
changes incorporate revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The proposed changes do not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient events.
The proposed change does not involve the addition or removal of any
equipment, or any design changes to the facility. Therefore, these
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.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4)
regarding the inservice testing of pumps and valves which are
classified as ASME Code Class 1, Class 2, and Class 3. The proposed
changes incorporate revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The proposed changes do not involve a modification to the physical
configuration of the plant nor does it involve a change in the methods
governing normal plant operation. The proposed changes will not impose
any new or different requirements or introduce a new accident
initiator, accident precursor, or malfunction mechanism. Additionally,
there is no change in the types or increases in the amounts of any
effluent that may be released offsite and there is no increase in
individual or cumulative occupational exposure. Therefore, the
[[Page 15770]]
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4)
regarding the inservice testing of pumps and valves which are
classified as ASME Code Class 1, Class 2, and Class 3. The proposed
changes do not involve a modification to the physical configuration of
the plant nor does it change the methods governing normal plant
operation. The proposed changes incorporate revisions to the ASME Code
that result in a net improvement in the measures for testing pumps and
valves. The safety function of the affected pumps and valves will be
maintained. 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: October 2, 2008.
Description of amendment request: The proposed amendments would
revise technical specifications (TS) associated with the verification
of ice condenser door operability. The proposed amendment affects the
current TS surveillance requirements 3.6.13.5 and 3.6.13.6.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (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 only analyzed accidents of possible consideration in regards to
changes potentially affecting the ice condenser are a loss of coolant
accident (LOCA) and a high energy line break (HELB) inside Containment.
However, the ice condenser is not postulated as being the initiator of
any LOCA or HELB. This is because it is designed to remain functional
following a design basis earthquake, and the ice condenser does not
interconnect or interact with any systems that interconnect or interact
with the Reactor Coolant or Main Steam Systems. Since these proposed
changes do not result in, or require, any physical change to the ice
condenser that could introduce an interaction with the Reactor Coolant
or Main Steam Systems, then there can be no change in the probability
of an accident previously evaluated. Regarding consequences of analyzed
accidents, the ice condenser is an engineered safety feature designed,
in part, to limit the Containment sub-compartment and Containment
vessel pressure immediately following the initiation of a LOCA or HELB.
Conservative sub-compartment and Containment pressure analysis shows
these criteria will be met if the total ice mass within the ice bed is
maintained in accordance with the DBA analysis; therefore, the proposed
TS [technical specification] SR [surveillance requirement] changes of
these requirements will not increase the consequences of any accident
previously evaluated.
Thus, based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
As previously described, the ice condenser is not postulated as
being the initiator of any design basis accident. The proposed changes
do not impact any plant system, structure or component that is an
accident initiator. The proposed TSs and TS Bases changes do not
involve any hardware changes to the ice condenser or other change that
could create any new accident mechanisms. Therefore, there can be no
new or different accidents created from those already identified and
evaluated.
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of the
fission product barriers to perform their design functions during and
following an accident situation. These barriers include the fuel
cladding, the reactor coolant system, and the Containment system. The
performance of the fuel cladding and the reactor coolant system will
not be impacted by the proposed changes. The Application provides a
description of additional sub-compartment and Containment pressure
response analysis that has been performed. This analysis demonstrates
that Containment will remain fully capable of performing its design
function with implementation of the proposed changes. Therefore, no
safety margin will be significantly impacted.
The changes proposed in this LAR [license amendment request] do not
make any physical alteration to the ice condenser doors, nor does it
affect the required functional capability of the doors in any way. The
intent of the proposed changes to the ice condenser door surveillance
requirements is to eliminate an unnecessary and overly restrictive
Lower Inlet Door torque surveillance test. There will be no degradation
in the operable status of the ice condenser doors and the ability to
confirm operability for the ice condenser doors will be maintained,
such that the doors will continue to fully perform their safety
function as assumed in the plant's safety analyses.
Thus, it can be concluded that the proposed TS and TS Bases changes
do not involve a significant reduction in the 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie Wong.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: February 24, 2009.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Surveillance Requirement (SR)
that governs operability testing of the pressure suppression chamber-
drywell vacuum breakers to incorporate the SR contained within the
Standard
[[Page 15771]]
Technical Specifications (STS), NUREG-1433 and delete the SR that
requires inspection of the pressure suppression chamber-drywell vacuum
breakers. Periodic inspections of the pressure suppression chamber-
drywell vacuum breakers are not required by the STS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The proposed amendment does not impact the operability of any
structure, system or component that affects the probability of an
accident or that supports mitigation of an accident previously
evaluated. The proposed amendment does not affect reactor operations or
accident analysis and has no radiological consequences. The operability
requirements for accident mitigation systems remain consistent with the
licensing and design basis. Therefore, the proposed amendment does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. The operation of VY in accordance with the proposed amendment
will not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed amendment does not change the design or function of
any component or system. No new modes of failure or initiating events
are being introduced. Therefore, operation of VY in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The operation of VY in accordance with the proposed amendment
will not involve a significant reduction in a margin of safety.
The proposed amendment does not change the design or function of
any component or system. The proposed amendment does not involve any
safety limits, safety settings or safety margins. The ability of the
pressure suppression chamber-drywell vacuum breakers to perform its
intended function will continue to be required in accordance with the
VY Technical Specifications.
Since the proposed controls are adequate to ensure the operability
of the pressure suppression chamber-drywell vacuum breakers, there will
still be high assurance that the components are operable and capable of
performing their respective functions. Therefore, operation of VY in
accordance with the proposed amendment will not involve a significant
reduction in the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: October 23, 2008.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to support the application of
alternative source term (AST) methodology with respect to the loss-of-
coolant accident and the fuel handling accident. The proposed request
is to support a full-scope application of an AST methodology, with the
exception that Technical Information Document (TID)-14844,
``Calculation of Distance Factors for Power and Test Reactor Sites,''
will continue to be used as the radiation dose basis for equipment
qualification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The implementation of AST assumptions has been evaluated in
revisions to the analyses of the following limiting design basis
accidents at LSCS [LaSalle County Station]:
Loss-of-Coolant Accident, and
Fuel Handling Accident
Based upon the results of these analyses, it has been demonstrated
that, with the requested changes, the dose consequences of these
limiting events are within the regulatory requirements and guidance
provided by the NRC for use with AST. The regulatory requirements and
guidance is presented in 10 CFR 50.67, ``Accident source term,'' and
associated NRC Regulatory Guide 1.183 and Standard Review Plan section
15.0.1. The AST is an input to calculations used to evaluate the
consequences of an accident, and does not by itself affect the plant
response, or the actual pathway of the radiation released from the
fuel. It does, however, better represent the physical characteristics
of the release, so that appropriate mitigation techniques may be
applied. Therefore, the consequences of an accident previously
evaluated are not significantly increased.
The equipment affected by the proposed change is mitigative in
nature, and relied upon after an accident has been initiated.
Application of the AST does not involve any physical changes to the TS,
while they revise certain performance requirements, do not involve any
physical modifications to the plant. As a result, the proposed changes
do not affect any of the parameters or conditions that could contribute
to the initiation of any accidents. As such, removal of operability
requirements during the specified conditions will not significantly
increase the probability of occurrence for an accident previously
analyzed. Since plant design basis accidents initiators are not being
altered by adoption of the AST analyses, the probability of an accident
previously evaluated is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be installed
and there are no physical modifications to existing equipment
associated with the proposed change). Similarly, it does not physically
change any structures, systems, or components involved in the
mitigation of any accidents. Thus, no new initiators or precursors of a
new or different kind of accident are created.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
[[Page 15772]]
Safety margins and analytical conservatisms have been evaluated and
have been found to be acceptable. The analyzed events have been
carefully selected and margin has been retained to ensure that the
analyses adequately bound postulated event scenarios. The dose
consequences due to design basis accidents comply with the requirements
of 10 CFR 50.67 and guidance of Regulatory Guide 1.183.
The proposed change is associated with the implementation of a new
licensing basis for LSCS design basis accidents. Approval of the change
from the original source term to a new source term taken from
Regulatory Guide 1.183 is being requested. The results of the accident
analyses, revised in support of the proposed license amendment, are
subject to revised acceptance criteria. The analyses have been
performed using conservative methodologies, as specified in Regulatory
Guide 1.183. Safety margins have been evaluated and analytical
conservatism has been utilized to ensure that the analyses adequately
bound the postulated limiting event scenario. The dose consequences of
these design basis accidents remain within the acceptance criteria
presented in 10 CFR 50.67 and Regulatory Guide 1.183.
The proposed change continues to ensure that the doses at the
exclusion area boundary and low population zone boundary, as well as
the control room, are within corresponding regulatory limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County,
Texas
Date of amendment request: February 11, 2009.
Brief description of amendment: The proposed amendment consists of
administrative revision to the operating licenses and Technical
Specifications (TSs) to revise the station name from Comanche Peak
Steam Electric Station (CPSES) to Comanche Peak Nuclear Power Plant
(CPNPP); remove the Table of Contents from TSs and maintain and revise
it in accordance with plant administrative procedures; delete TSs
3.2.1.1, 3.2.3.1, 5.5.9.1, 5.6.10 and several footnotes from Tables
3.3.1-1, 3.3.2-1, and TS 3.4.10 since these TSs and footnotes are no
longer applicable to CPSES, Units 1 and 2 operation; delete several
topical reports from the list of approved analytical methods used to
determine core operating limits in TS 5.6.5, no longer in use, since
these topical reports have been replaced by standard Westinghouse
methods and Westinghouse methods have been approved for use at CPSES,
Units 1 and 2, under a separate amendment request; make editorial
corrections; and reprint and reissue the entire TS due to adoption of
`FrameMaker' software in place of `Microsoft Word' software.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the station name, removes the Table of
Contents from the Technical Specifications, deletes several Technical
Specifications and footnotes which are no longer applicable to [CPSES]
Unit 1 or Unit 2 operation, renumbers subsequent Technical
Specifications, deletes several topical reports from the list of
approved analytical methods used to determine core operating limits,
and corrects various editorial and formatting errors. The Table of
Contents does not include information required by 10 CFR 50.36 [Title
10 of the Code of Federal Regulations, Section 50.36] to be reviewed by
the NRC [U.S. Nuclear Regulatory Commission] staff and is not required
by the regulation. The Technical Specifications and footnotes which are
being deleted were only applicable during previous operational cycles
and are now defunct requirements since both Units have completed the
applicable operational cycles. The topical reports deleted from
Technical Specification 5.6.5b are no longer used to determine the core
operating limits for Comanche Peak Nuclear Power Plant. The remaining
topical reports listed in Technical Specification 5.6.5b will be used
to determine the core operating limits for both Comanche Peak Nuclear
Power Plant units. All other changes proposed are corrections of
previous inadvertent editorial errors or changes in format to increase
conformity with the guidelines described in TSTF-RPT-01, ``Writer's
Guide for Plant-Specific Improved Technical Specifications'', published
in June, 2005. All of the proposed changes are administrative changes
which do not change the meaning, intent, interpretation, or application
of the Technical Specifications. None of the proposed changes affect
the operation, physical configuration, or function of plant equipment
or systems. The changes do not affect the initiators or assumptions of
analyzed events; nor do they impact the mitigation of accidents or
transient events. 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 change revises the station name, removes the Table of
Contents from the Technical Specifications, deletes several Technical
Specifications and footnotes which are no longer applicable to [CPSES,]
Unit 1 or Unit 2 operation, renumbers subsequent Technical
Specifications, deletes several topical reports from the list of
approved analytical methods used to determine core operating limits,
and corrects various editorial and formatting errors. The Table of
Contents does not include information required by 10 CFR 50.36 to be
reviewed by the Nuclear Regulatory Commission staff and is not required
by the regulation. The Technical Specifications and footnotes which are
being deleted were only applicable during previous operational cycles
and are now defunct requirements since both Units have completed the
applicable operational cycles. The topical reports deleted from
Technical Specification 5.6.5b are no longer used to determine the core
operating limits for Comanche Peak Nuclear Power Plant. The remaining
topical reports listed in Technical Specification 5.6.5b will be used
to determine the core operating limits for both Comanche Peak Nuclear
Power Plant units. All other changes proposed are corrections of
previous inadvertent editorial errors or changes in format to increase
conformity with the guidelines described in TSTF-RPT-01, ``Writer's
Guide for Plant-Specific Improved Technical Specifications'', published
in June, 2005. All of the proposed changes are administrative changes
which do not change the meaning, intent,
[[Page 15773]]
interpretation, or application of the Technical Specifications. None of
the changes alter the plant configuration, require installation of new
equipment, alter assumptions about previously analyzed accidents, or
impact the operation or function of any plant equipment or systems.
Therefore, the proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the station name, removes the Table of
Contents from the Technical Specifications, deletes several Technical
Specifications and footnotes which are no longer applicable to [CPSES,]
Unit 1 or Unit 2 operation, renumbers subsequent Technical
Specifications, deletes several topical reports from the list of
approved analytical methods used to determine core operating limits,
and corrects various editorial and formatting errors. The Table of
Contents does not include information required by 10 CFR 50.36 to be
reviewed by the Nuclear Regulatory Commission staff and is not required
by the regulation. The Technical Specifications and footnotes which are
being deleted were only applicable during previous operational cycles
and are now defunct requirements since both Units have completed the
applicable operational cycles. The topical reports deleted from
Technical Specification 5.6.5b are no longer used to determine the core
operating limits for Comanche Peak Nuclear Power Plant. The remaining
topical reports listed in Technical Specification 5.6.5b will be used
to determine the core operating limits for both Comanche Peak Nuclear
Power Plant units. All other changes proposed are corrections of
previous inadvertent editorial errors or changes in format to increase
conformity with the guidelines described in TSTF-RPT-01, ``Writer's
Guide for Plant-Specific Improved Technical Specifications'', published
in June, 2005. All of the proposed changes are administrative changes
which do not change the meaning, intent, interpretation, or application
of the Technical Specifications. None of the proposed changes alter the
effective technical content of the Technical Specifications. Therefore
the proposed changes do not involve a 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: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 31, 2008.
Description of amendment request: The proposed amendment modifies
the surveillance requirements in Technical Specification (TS) 3.6(3),
``Containment Recirculating Air Cooling and Filtering System,'' and
removes the license conditions related to the replacement and testing
of containment air cooling and filtering (CACF) unit high-efficiency
particulate air (HEPA) filters and surveillance testing of the CACF
unit relief ports. These license conditions were committed to by the
licensee in its letter dated April 10, 2008 (Agencywide Documents
Access and Management System (ADAMS) Accession No. ML081010122), and
implemented via TS Amendment No. 255 (ADAMS Accession No. ML081140390),
dated May 2, 2008.
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 containment air cooling and filtering system (CACFS) is not an
initiator of any accident previously evaluated at the Fort Calhoun
Station (FCS). The CACFS is an accident mitigation system. The design
basis function of the CACFS is to limit the containment pressure rise
by providing a means for cooling the containment following a loss-of-
coolant accident (LOCA) or main steam line break (MSLB). In accordance
with TS Amendment No. 255, the CACFS high efficiency particulate air
(HEPA) filters are also credited to reduce post-LOCA radioactive
leakage from containment.
The proposed changes provide additional assurance that the CACFS is
capable of performing its design and licensing basis functions to
mitigate these design basis accidents (DBAs). The CACFS face and bypass
dampers are aligned to their accident positions permanently causing the
CACFS to operate in filtered air mode. Surveillance testing has shown
that operating the system in this alignment over long periods does not
jeopardize filter performance. Over the lifetime of the plant, the
differential pressures measured across the combined HEPA and charcoal
filter banks have met test acceptance criteria.
Increasing the number of surveillance requirements will not
adversely affect the function of the CACFS but rather provides
additional assurance that the CACFS is capable of responding to a DBA.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The CACFS was designed to remove heat released to the containment
atmosphere during a DBA to the extent necessary to maintain the
containment structure below its design pressure. The containment
airflow continually passes through the cooling coils. The proposed
changes to the surveillance requirements do not affect the active
function of the CACFS.
The CACFS will continue to operate in normal and accident
conditions to remove heat and radioactive particulates and aerosols.
The proposed changes enhance surveillance testing of the CACFS by
requiring more frequent exercising of the fans, imposing a more
stringent pressure drop limit, specifying a HEPA filter replacement
interval, and instituting a requirement to exercise the relief ports.
These changes ensure that the CACFS is capable of long-term operation
in filtered air mode while remaining capable of providing cooling and
filtering sufficient to mitigate design basis accidents.
No credible new failure mechanisms, malfunctions, or accident
initiators not previously considered in the design and licensing basis
are created and none of the initial condition assumptions of any
accident evaluated in the safety analysis are impacted.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The containment building and associated penetrations are designed
to
[[Page 15774]]
withstand an internal pressure of 60 psig [pounds per square inch
gauge] at 305[deg]F, including all thermal loads resulting from the
temperature associated with this pressure, with a leakage rate of 0.1
percent by weight or less of the contained volume per 24 hours. [Omaha
Public Power District] credits the CACFS in the containment pressure
analysis for a LOCA, and for the containment pressure response to a
main steam line break (MSLB).
The proposed changes impose more stringent surveillance test
requirements. This provides additional assurance that the CACFS will
perform its design basis and licensing basis functions to be capable of
long-term post-DBA operation in filtered air mode to limit the
containment temperature and pressure increase to within design limits
and to reduce post-LOCA radioactive leakage from containment.
Neither the design basis nor the licensing basis for post-DBA
containment heat removal is adversely affected by the proposed changes.
The ability to maintain design limits for containment peak pressure and
temperature, as well as long-term containment pressure and temperature,
are preserved.
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: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 30, 2009.
Description of amendment request: The proposed amendment would
modify the Fort Calhoun Station (FCS), Unit No. 1, Renewed Operating
License No. DPR-40, by adding operability and surveillance testing
requirements to the FCS Technical Specifications (TS) for the steam
generator (SG) blowdown isolation on a reactor trip. Specifically, the
proposed changes will revise TS Limiting Conditions for Operation (LCO)
2.15, Instrumentation and Control Systems, Table 2-4, Instrument
Operating Conditions for Isolation Functions, to include operability
requirements for SG blowdown isolation on a reactor trip and to add
applicable footnotes. In addition, TS 3.1, Instrumentation and Control,
Table 3-2, Minimum Frequencies for Checks, Calibrations and Testing of
Engineered Safety Features, Instrumentation and Controls, is being
revised to include the surveillance test requirements for SG blowdown
isolation on a reactor trip. An administrative change is also being
made to TS LCO 2.15(1), to delete the words ``key operated'' as the
``key'' associated with the bypass switches is not a critical element
in controlling the use of bypass switches. This amendment will allow
FCS to credit an automatic SG blowdown isolation interlock being
installed during the 2009 Refueling Outage (RFO).
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 provides Technical Specification (TS)
operability and surveillance testing requirements for automatic steam
generator (SG) blowdown isolation on a reactor trip in the event of a
loss of main feedwater (LMFW). Automatic isolation will ensure that the
existing 15-minute requirement in the Updated Safety Analysis Report
(USAR) Chapter 14.10 safety analysis is met without the risk that an
unanticipated distraction could prevent manual action from occurring at
the proper time. The installation of this feature will eliminate the
need for manual isolation of blowdown and thus will eliminate the
associated operator challenge.
Automatic isolation of blowdown will reduce the consequences of the
LMFW event by providing automatic isolation prior to manual isolation
being initiated by the operators. Automatic isolation at the time of
reactor trip will reduce the severity of the LMFW event by isolating
the SGs earlier in the event, thereby conserving SG inventory.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new malfunctions are being introduced by this activity, and
based on the current redundancy in the design, there are no
malfunctions of the SG blowdown isolation valves that challenge nuclear
safety.
The SG blowdown isolation valves will continue to function as
currently credited for the LMFW event; thus, this proposed change does
not alter their ability to function as containment isolation valves to
maintain containment integrity. The manual isolation capability remains
unchanged.
A failure analysis has been prepared which shows that the addition
of the automatic isolation feature does not introduce a new failure
mode or malfunction to the valve circuits. An isolation of SG blowdown,
either through the designed circuit following a reactor trip, or during
normal operations, does not present a nuclear safety challenge. The
capability exists for operators to bypass the isolation signal and
restore blowdown as plant conditions warrant.
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 addition of an automatic isolation interlock to the SG blowdown
isolation valve circuits that close the valves on a reactor trip
actually increases the margin of safety by isolating the SG early in
the event to maintain SG inventories.
A reactor trip signal is generated in the first seconds of an LMFW
due to reduced SG inventories. Because it is desirable to isolate
blowdown as soon as possible following the LMFW event, for maximum
margin, a reactor trip signal will be used for the SG blowdown
isolation interlock. Isolating blowdown earlier in an event provides
greater operating margin in terms of maximizing SG inventories. More
margin allows operators more time to address operator demands that
occur during transient events.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 15775]]
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 30, 2009.
Description of amendment request: The proposed amendment would
delete those portions of the Technical Specifications (TS) superseded
by Title 10 of the Code of Federal Regulations (10 CFR) Part 26,
Subpart I. The licensee is proposing to adopt the approved Technical
Specification Task Force (TSTF) change traveler TSTF-511, Revision 0,
``Eliminate Working Hour Restrictions from TS 5.2.2 to Support
Compliance with 10 CFR Part 26.''
The NRC staff issued a ``Notice of Availability of Model Safety
Evaluation, Model No Significant Hazards Determination, and Model
Application for Licensees That Wish To Adopt TSTF-511, Revision 0,
``Eliminate Working Hour Restrictions From TS 5.2.2 To Support
Compliance With 10 CFR Part 26,'' in the Federal Register on December
30, 2008 (73 FR 79923). The notice included a model safety evaluation,
a model no significant hazards consideration (NSHC) determination, and
a model license amendment request, using the consolidated line item
improvement process. In its application dated January 30, 2009, the
licensee affirmed the applicability of the model NSHC determination,
which is presented below.
Basis for proposed (NSHC) determination: As required by 10 CFR
50.91(a), an analysis of the issue of NSHC determination 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 removes Technical Specification restrictions on
working hours for personnel who perform safety related functions. The
Technical Specification restrictions are superseded by the worker
fatigue requirements in 10 CFR Part 26. Removal of the Technical
Specification requirements will be performed concurrently with the
implementation of the 10 CFR Part 26, Subpart I, requirements. The
proposed change does not impact the physical configuration or function
of plant structures, systems, or components (SSCs) or the manner in
which SSCs are operated, maintained, modified, tested, or inspected.
Worker fatigue is not an initiator of any accident previously
evaluated. Worker fatigue is not an assumption in the consequence
mitigation of any accident previously evaluated. Therefore, it is
concluded that this 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 removes Technical Specification restrictions on
working hours for personnel who perform safety related functions. The
Technical Specification restrictions are superseded by the worker
fatigue requirements in 10 CFR Part 26. Working hours will continue to
be controlled in accordance with NRC requirements. The new rule allows
for deviations from controls to mitigate or prevent a condition adverse
to safety or as necessary to maintain the security of the facility.
This ensures that the new rule will not unnecessarily restrict working
hours and thereby create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed change
does not alter the plant configuration, require new plant equipment to
be installed, alter accident analysis assumptions, add any initiators,
or effect the function of plant systems or the manner in which systems
are operated, maintained, modified, tested, or inspected. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change removes Technical Specification restrictions on
working hours for personnel who perform safety related functions. The
Technical Specification restrictions are superseded by the worker
fatigue requirements in 10 CFR Part 26. The proposed change does not
involve any physical changes to plant or alter the manner in which
plant systems are operated, maintained, modified, tested, or inspected.
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
affected by this change. The proposed change will not result in plant
operation in a configuration outside the design basis. The proposed
change does not adversely affect systems that respond to safely
shutdown the plant and to maintain the plant in a safe shutdown
condition. Removal of plant-specific Technical Specification
administrative requirements will not reduce a margin of safety because
the requirements in 10 CFR Part 26 are adequate to ensure that worker
fatigue is managed. 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 amendment request involves NSHC.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: December 19, 2008.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) to (1) correct an error in TS
Table 3.3.2-1, ``Engineered Safety Feature Actuation System
Instrumentation,'' Function 1.a, to reflect the correct CONDITIONS for
applicable Modes 1, 2, 3, and 4, (2) revise TS Limiting Condition for
Operation (LCO) 3.3.4 degraded voltage relay and loss of voltage relay
Limiting Safety System Settings values to reflect the revised analysis,
and (3) revise the load requirement of Surveillance Requirement 3.8.1.3
to reflect values supported by the diesel generator accident loading
analyses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to LCO 3.3.2 correct an administrative error
which directed inadequate action in the event that a channel of
instrumentation is lost for manual safety injection initiation. The
amendment places the plant in a
[[Page 15776]]
more conservative condition, Mode 5, if the other Required Actions
cannot be executed within their periodicity.
The proposed changes to LCO 3.3.4 provide setpoint changes based on
a revised calculation, which generated new setpoints for the loss of
voltage relays and degraded voltage relays. The new setpoints ensure
the protective relays will function when required, will ensure
protection from thermal damage to loads on the 480V busses, and will
not cause unintended diesel generator starts even in worst case
scenarios, with power provided from offsite.
The proposed changes to LCO 3.8.1 involve an increase in the
minimum load band value for diesel generator surveillance SR 3.8.1.3.
This change ensures that the diesel generators are capable of
synchronizing with the offsite electrical system and accepting loads
greater than or equal [to] the equivalent of the maximum expected
accident loads. The new load band value is more conservative than the
existing value and provides a more thorough test to ensure equipment
emergency response capability.
Therefore, the probability or consequences of an accident
previously evaluated will not be significantly increased.
2. Do the proposed amendments create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes involve correcting an administrative error and
revising previously established values associated with the diesel
generators to increase conservatism. None of these proposed changes
involve a physical alteration of the plant (i.e., no new or different
types of equipment will be installed) or a change in methods governing
normal plant operation. The proposed changes preserve the safety
analysis assumptions related to accident mitigation. No initiators or
accident precursors are created by this change. Therefore, the
possibility of a new or different kind of accident not previously
evaluated is not created.
3. Do the proposed amendments involve a significant reduction in a
margin of safety?
Response: No.
The level of safety of facility operation is unaffected by any of
the proposed changes. The requested administrative change is
conservative compared to the existing requirement. The response of the
diesel generators to accident transients reported in the Updated Final
Safety Analysis Report (UFSAR) is unaffected by these changes. The
proposed changes preserve the safety analysis assumptions related to
accident mitigation. Therefore, these 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: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor,
Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, 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].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 7, 2008, as supplemented
by letters dated December 17, 2008, and March 9, 2009.
Brief description of amendments: The amendments revise Surveillance
Requirement (SR) 3.6.1.6.1 to add a new requirement to verify that each
vacuum breaker is closed within 6 hours following an operation that
causes any of the vacuum breakers to open and, also, revise SR
3.6.1.6.2 by removing the requirement to perform functional testing of
each vacuum breaker within 12 hours following an operation that causes
any of the vacuum breakers to open.
Date of issuance: March 11, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 251 and 279.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: September 23, 2008 (73
FR 54864). The supplemental letter provided clarifying information that
was within the scope of the initial notice 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 March 11, 2009.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
[[Page 15777]]
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353,
Limerick Generating Station, Unit 1 and 2, Montgomery County,
Pennsylvania
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendments: April 21, 2008, as supplemented
on March 11, 2009.
Brief description of amendments: The proposed amendment removes
references to and limits provided by Nuclear Regulatory Commission
Generic Letter (GL) 82-12, ``Nuclear Power Plant Staff Working Hours,''
from the subject plants' technical specifications (TS). The references
and limitations have been superseded by the requirements of Title 10 of
the Code of Federal Regulations, Part 26 (10 CFR 26), Subpart I,
``Managing Fatigue.''
Date of issuance: March 23, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment Nos.: 157, 157, 162, 162, 185, 231, 224, 192, 179, 198,
159, 274, 271, 275, 243, 238, 270.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66,
NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, NPF-39, NPF-85, DPR-16, DPR-44,
DPR-56, DPR-29, DPR-30, DPR-50: The amendments revised the Technical
Specifications/Licenses.
Date of initial notice in Federal Register: June 3, 2008 (73 FR
31721). The March 11, 2009, supplement contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 23, 2009.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 24, 2008, as supplemented by
letters dated September 11 and 19, 2008, November 6, 2008, and February
26, 2009.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 3.7.3, ``Reactor Equipment Cooling (REC)
System,'' to allow credit for the ability to align the service water
system to the REC system.
Date of issuance: March 20, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 232.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 22, 2008 (73 FR
21660). The supplemental letters dated September 11 and 19, 2008,
November 6, 2008, and February 26, 2009, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 20, 2009.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York
Date of application for amendment: August 15, 2008, as supplemented
on December 4, 2008.
Brief description of amendments: The amendment revises NMP1
Technical Specification (TS) 6.5.7, ``10 CFR 50 [Part 50 of Title 10 of
the Code of Federal Regulations Appendix J Testing Program Plan,'' to
allow a one-time extension of the Integrated Leak Rate Test (ILRT)
interval for no more than 5 years. The amendment allows the next ILRT
for NMP1 to be performed within 15 years from the last ILRT as opposed
to the current 10-year interval.
Date of issuance: March 11, 2009.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 202.
Renewed Facility Operating License No. DPR-063: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: October 21, 2008 (73 FR
62566). The supplement dated December 4, 2008, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 11, 2009.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, LLC, Docket No. 50-263,
Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of application for amendment: April 3, 2008, as supplemented
on February 23, 2009.
Brief description of amendment: The amendment adopted the proposed
requirements regarding control room envelope habitability set forth in
Technical Specifications Task Force (TSTF) change traveler TSTF-448,
Revision 3. Specifically, the amendment revised the requirements in TS
Section 3.7.4, ``Control Room Emergency Filtration (CREF) System,''
adds a new TS Section 5.5.13, ``Control Room Envelope Habitability
Program,'' and added a license condition to the operating license to
implement the TS changes.
Date of issuance: March 17, 2009.
Effective date: As of the date of issuance and shall be implemented
by November 1, 2009.
Amendment No.: 160.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 6, 2008 (73 FR
25043). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 17, 2009.
No significant hazards consideration comments received: No.
[[Page 15778]]
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: July 31, 2008.
Brief description of amendments: The amendments changed the PPL
Susquehanna, LLC (PPL) Units 1 and 2 Technical Specification 3.6.1.3
``Primary Containment Isolation Valves (PCIVs).'' It revised the
Secondary Containment Bypass Leakage limit in Surveillance Requirement
3.6.1.3.11 from ``less than or equal to 9 standard cubic foot/feet per
hour (scfh)'' to ``less than or equal to 15 scfh when pressurized to
greater than or equal to Pa.''
Date of issuance: March 18, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 251 for Unit 1 and 231 for Unit 2.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: November 18, 2008 (73
FR 68455). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation (SE) dated March 18, 2009.
No significant hazards consideration comments received: No.
However, comments have been received from the Commonwealth of
Pennsylvania and have been addressed in the SE.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: March 19, 2008, as supplemented
October 7, 2008, November 17, 2008, and December 10, 2008.
Brief description of amendment: The amendments revise the technical
specifications (TSs) to (1) delete TS 3.7.13, ``MCR/ESGR Bottled Air
System,'' (2) create TS 3.3.6, ``Main Control Room/Emergency Switchgear
Room (MCR/ESGR) Envelope Isolation Actuation Instrumentation,'' to
establish the operability requirements for the MCR/ESGR envelope
isolation function, and (3) incorporate TS 3.7.14, ``MCR/ESGR Emergency
Ventilation During Movement of Recently Irradiated Fuel Assemblies,''
into TS 3.7.10, ``MCR/ESGR Emergency Ventilation System.'' The changes
revise the TSs to be consistent with the assumptions of the current
dose analysis of record, performed in accordance with Title 10 of the
Code of Federal Regulations, Section 50.67, ``Accident Source Term,''
and the results of the nonpressurized MCR/ESGR envelope tracer gas
testing.
Date of issuance: March 25, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 255/236.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: April 22, 2008 (73 FR
21661). The supplements dated October 7, 2008, November 17, 2008, and
December 10, 2008, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated March 25, 2009.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th of March, 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-7494 Filed 4-6-09; 8:45 am]
BILLING CODE 7590-01-P