[Federal Register Volume 74, Number 163 (Tuesday, August 25, 2009)]
[Notices]
[Pages 42926-42940]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-20403]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0363]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires the Commission publish notice of any amendments issued, or
proposed to be issued and grants the Commission the authority to issue
and make immediately effective any amendment to an operating license
upon a determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 30, 2009 to August 12, 2009. The last
biweekly notice was published on August 11, 2009 (74 FR 40233).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the 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 (RDB), TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or
copied for a fee, at the NRC's Public Document Room (PDR), located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the 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
[[Page 42927]]
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, 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 2007 (72 FR 49139, August 28, 2007). The E-Filing
process requires participants to submit and serve all adjudicatory
documents over the Internet, or in some cases to mail copies on
electronic storage media. Participants may not submit paper copies of
their filings unless they seek an exemption in accordance with the
procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the petitioner/requestor
should 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 ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM 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 using the agency's adjudicatory e-
filing system 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 toll-free help line
number is 1-866-672-7640. A person filing electronically may also seek
assistance by sending an e-mail to the NRC electronic filing Help Desk
at [email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the request and/
or petition 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 submissions.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly
[[Page 42928]]
available records will be accessible from the ADAMS Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who
encounter problems in accessing the documents located in ADAMS, should
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737, or
by e-mail to [email protected].
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: June 29, 2009.
Description of amendment request: The proposed amendment would
revise the requirements in Technical Specification (TS) 5.5.6,
``Inservice Testing Program.'' TS 5.5.6 currently contains references
to the American Society of Mechanical Engineers Boiler and Pressure
Vessel Code (ASME Code), Section XI as the source of requirements for
the inservice testing (IST) of ASME Code Class 1, 2, and 3 pumps and
valves. The proposed changes would delete the references to Section Xl
of the ASME Code and incorporate references to the ASME Code for
Operation and Maintenance of Nuclear Power Plants (ASME OM Code). In
addition, the proposed amendment would address the applicability of
Surveillance Requirement 3.0.2 to other normal and accelerated
frequencies as 2 years or less in the IST program. These changes are
consistent with changes identified in the Improved Standard Technical
Specifications (ISTS) by Technical Specification Task Force Traveler
(TSTF) Nos. 479 and 497.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Technical Specification
Inservice Testing Program for consistency with the requirements of
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. The proposed change incorporates revisions to the
ASME Code that result in a net improvement in the measures for
testing pumps and valves.
The proposed changes revise TS 5.5.6 for RBS to conform to the
requirements of 10 CFR 50.55a(f) regarding the IST of pumps and
valves for the third 10-Year Interval. The current TS reference the
ASME Boiler and Pressure Vessel Code, Section XI, requirements for
the IST of ASME Code Class 1, 2, and 3 pumps and valves. The
proposed changes would reference the ASME OM Code instead. This is
consistent with 10 CFR 50.55a(f). The proposed changes are
administrative in nature.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. They do not involve the addition or removal of any
equipment, or any design changes to the facility.
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 revises the Technical Specification
Inservice Testing Program for consistency with the requirements of
10 CFR 50.55a(f)(4) for pumps and valves which are classified as
ASME Code Class 1, Class 2 and Class 3. The proposed change
incorporates revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The proposed TS changes do not involve physical changes to the
facility. In addition, the proposed changes have no affect on plant
configuration, or method of operation of plant structures, systems,
or components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the Technical Specification
Inservice Testing Program for consistency with the requirements of
10 CFR 50.55a(f)(4) for pumps and valves which are classified as
ASME Code Class 1, Class 2 and Class 3. The proposed change
incorporates revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The change does not involve a physical change to the plant or a
change in the manner in which the plant is operated or controlled.
The IST of the Class 1, 2, and 3 pumps and valves continue to meet
the appropriate requirements.
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: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: July 3, 2009.
Description of amendment request: The proposed amendments would
revise the operability requirements and actions in Technical
Specification (TS) 3.4.15, ``RCS [Reactor Coolant System] Leakage
Detection Instrumentation,'' and the associated Bases Section to
reflect the revised TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change reduces the time allowed for the plant to
operate when the only Technical Specification (TS) 3.4.15 operable
Reactor Coolant System (RCS) leakage instrumentation monitor is the
containment atmosphere gaseous radioactivity monitor, and revises
the basis for operability for the containment sump monitors,
containment atmosphere particulate radioactivity monitor,
containment atmosphere gaseous radioactivity monitor, and the
containment fan cooler unit condensate collection monitor. The
proposed change increases the allowed operating time when all RCS
leakage detection system instrumentation is inoperable. The proposed
change also removes the word ``required'' from TS 3.4.15 Condition
A, Required Action A.2, Condition B, and Required Action B.2,
revises TS 3.4.15 Condition A to apply to any containment sump
monitor, and revises the name of the containment fan cooler unit
(CFCU) condensate collection monitor in the TS 3.4.15 Actions. The
monitoring of RCS leakage is not a precursor to any accident
previously evaluated. The monitoring of RCS leakage is not used to
mitigate the consequences of any accident previously evaluated.
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 accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or the addition of new or different type of
[[Page 42929]]
equipment. The change does not involve a change in how the plant is
operated.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The change that reduces the allowed time of operation with only
the least accurate containment atmosphere gaseous radiation monitor
operable increases the margin of safety by increasing the likelihood
that an increase in RCS leakage will be detected before it
potentially results in gross failure. For the change that allows a
limited period of time to restore at least one RCS leakage detection
monitor to operable status when all leakage detection monitors are
inoperable, two sources of diverse leakage detection capability are
required to be provided during the limited period. Allowing a
limited period of time to restore at least one RCS leakage detection
instrument to operable status before requiring a plant shutdown
avoids the situation of putting the plant through a thermal
transient without RCS leakage monitoring. The change to TS 3.4.15
Condition A, Required Action A.2, Condition B, Required Action B.2,
Condition C, and Required Action C.2.2 is consistent with TS
[Limiting Condition for Operation] 3.4.15 and does not impact the
RCS leakage instrumentation. The revision to the TS bases for
operability of the RCS leakage instrumentation monitors does not
involve a change in the leakage instrumentation and is consistent
with the original design of the leakage instrumentation.
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 requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Michael T. Markley.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: April 9, 2009.
Description of amendment request: The proposed amendment would
relocate Technical Specification (TS) requirements pertaining to
communications during refueling operations (TS 3/4.9.5), manipulator
crane operability (TS 3/4.9.6), and crane travel (TS 3/4.9.7) to the
Technical Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review is
presented below.
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would relocate TS requirements to the
Technical Requirements Manual (TRM) which is a licensee-controlled
document. The TS requirements to be relocated relate to control room
communications during refueling, operability of the manipulator
crane and auxiliary hoist for movement of control rods or fuel
assemblies within the reactor pressure vessel, and control of heavy
loads over fuel assemblies in the fuel storage pool. Once relocated,
any future changes would be controlled by 10 CFR 50.59. The proposed
amendment is administrative in nature from the standpoint that the
current TS requirements would be relocated verbatim to the TRM.
There are no physical plant modifications associated with this
change. The proposed amendment would not alter the way any
structure, system, or component (SSC) functions and would not alter
the way the plant is operated. As such, the proposed amendment would
have no impact on the ability of the affected SSCs to either
preclude or mitigate an accident. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment would not change the design function or
operation of the SSCs involved and would not impact the way the
plant is operated. As such, the proposed change would not introduce
any new failure mechanisms, malfunctions, or accident initiators not
already considered in the design and licensing bases. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is associated with the confidence in the
ability of the fission product barriers (i.e., fuel cladding,
reactor coolant pressure boundary, and containment structure) to
limit the level of radiation to the public. There are no physical
plant modifications associated with the proposed amendment. The
proposed amendment would not alter the way any SSC functions and
would not alter the way the plant is operated. The proposed
amendment would not introduce any new uncertainties or change any
existing uncertainties associated with any safety limit. The
proposed amendment would have no impact on the structural integrity
of the fuel cladding, reactor coolant pressure boundary, or
containment structure. Based on the above considerations, the NRC
staff concludes that the proposed amendment would not degrade the
confidence in the ability of the fission product barriers to limit
the level of radiation to the public. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 3, 2009.
Description of amendment request: The proposed amendment would
revise the Operating Licenses to deviate from certain South Texas
Project Fire Protection Program requirements. The amendment will allow
the performance of operator manual actions to achieve and maintain safe
shutdown in the event of a fire in lieu of meeting circuit separation
protection requirements of Title 10 of the Code of Federal Regulations
(10 CFR), Part 50, Appendix R, Section III.G.2 for Fire Area 31.
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 design functions of structures, systems and component[s] are
not impacted by the proposed change. The proposed change involves
operator manual actions in response to a fire and will not initiate
an event. The proposed actions do not increase the probability of
occurrence of a fire or any other accident previously evaluated.
The proposed actions are feasible and reliable and demonstrate
that the unit can be safely shutdown in the event of a fire. No
significant consequences result from the performance of the proposed
actions.
Therefore, the proposed change does not involve a significant
increase in the
[[Page 42930]]
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 design functions of structures, systems and component[s] are
not impacted by the proposed amendment. The proposed change involves
operator manual actions in response to a fire. They do not involve
new failure mechanisms or malfunctions that can initiate a new
accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Adequate time is available to perform the proposed operator
manual actions to account for uncertainties in estimates of the time
available and in estimates of how long it takes to diagnose and
execute the actions. The actions are straightforward and do not
create any significant concerns. The actions have been verified that
they can be performed through demonstration and they are
proceduralized. The proposed actions are feasible and reliable and
demonstrate that the unit can be safely shutdown in the event of a
fire.
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendment involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 3, 2009.
Description of amendment request: The proposed change would revise
the Operating Licenses to deviate from certain South Texas Project Fire
Protection Program requirements. The amendment will allow the
performance of operator manual actions to achieve and maintain safe
shutdown in the event of a fire in lieu of meeting circuit separation
protection requirements of Title 10 of the Code of Federal Regulations
(10 CFR), Part 50, Appendix R, Section III.G.2 for Fire Area 27.
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 design functions of structures, systems and components are
not impacted by the proposed change. The proposed change involves
operator manual actions in response to a fire, and will not initiate
an event. The proposed actions do not increase the probability of
occurrence of a fire or any other accident previously evaluated.
The proposed actions are feasible and reliable and demonstrate
that the unit can be safely shutdown in the event of a fire. No
significant consequences result from the performance of the proposed
actions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The design functions of structures, systems and components are
not impacted by the proposed amendment. The proposed change involves
operator manual actions in response to a fire. They do not involve
new failure mechanisms or malfunctions that can initiate a new
accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant rendition
in a margin of safety?
Response: No.
Adequate time is available to perform the proposed operator
manual actions to account for uncertainties in estimates of the time
available and in estimates of how long it takes to diagnose and
execute the actions. The actions are straightforward and do not
create any significant concerns. The actions have been verified that
they can be performed through demonstration and they are
proceduralized. The proposed actions are feasible and reliable and
demonstrate that the unit can be safely shutdown in the event of a
fare.
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendment involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: July 9, 2009.
Description of amendment request: The proposed amendment would
allow use of a dedicated on-line core power distribution monitoring
system (PDMS) to enhance surveillance of core thermal limits and would
revise Technical Specification (TS) TS 1.1, ``Definitions,'' TS 3.1.8,
``Rod Position Indication,'' TS 3.2.1, ``Heat Flux Hot Channel
Factor,'' TS 3.2.4, ``Quadrant Power Tilt Ratio (QPTR)'', and TS 3.3.1,
``Reactor Trip System (RTS) Instrumentation.''
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 Power Distribution Monitoring System (PDMS) performs
essentially continuous core power distribution monitoring with data
input from existing plant instrumentation. This system utilizes an
NRC-approved Westinghouse proprietary computer code, i.e. Best
Estimate Analyzer for Core Operations--Nuclear (BEACON), to provide
data reduction for incore flux maps, core parameter analysis, load
follow, operation simulation, and core prediction. The PDMS does not
provide any protection or control system function. Fission product
barriers are not impacted by these proposed changes. The proposed
changes occurring with PDMS will not result in any additional
challenges to plant equipment that could increase the probability of
any previously evaluated accident. The changes associated with the
PDMS do not affect plant systems such that their function in the
control of radiological consequences is adversely affected. These
proposed changes will, therefore, not affect the mitigation of the
radiological consequences of any accident described in the Updated
Final Safety Analysis Report (UFSAR).
Use of the PDMS supports maintaining the core power distribution
within required limits. Further, continuous on-line monitoring
through the use of PDMS provides significantly more information
about the power distributions present in the core than is currently
available. This result in more time (i.e. earlier determination of
an adverse condition developing) for operator action prior to having
an adverse condition develop that could lead to an accident
condition or to unfavorable initial conditions for an accident.
Therefore, the proposed change does not involve a significant
increase in the
[[Page 42931]]
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.
Other than use of the PDMS to monitor core power distribution,
implementation of the PDMS and associated Technical Specification
changes has no impact on plant operations or safety, nor does it
contribute in any way to the probability or consequences of an
accident. No safety related equipment, safety function, or plant
operation will be altered as a result of this proposed change. The
possibility for a new or different type of accident from any
accident previously evaluated is not created since the changes
associated with implementation of the PDMS do not result in a change
to the design basis of any plant component or system. The evaluation
of the effects of using the PDMS to monitor core power distribution
parameters shows that all design standards and applicable safety
criteria limits are met.
The proposed changes do not result in any event previously
deemed incredible being made credible. Implementation of the PDMS
will not result in any additional adverse condition and will not
result in any increase in the challenges to safety systems. The
cycle specific variables required by the PDMS are calculated using
NRC approved methods. The Technical Specifications will continue to
require operation within the required core operating limits, and
appropriate actions will continue to be taken when or if limits are
exceeded.
Therefore, the proposed change does not create the possibility
of a new or different kind of an accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
No margin of safety is adversely affected by the implementation
of the PDMS. The margins of safety provided by current Technical
Specification requirements and limits remain unchanged, as the
Technical Specifications will continue to require operation within
the core limits that are based on NRC approved reload design
methodologies. Appropriate measures exist to control the values of
these cycle specific limits, and appropriate actions will continue
to be specified and taken for when limits are violated. Such actions
remain unchanged.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: March 20, 2009.
Description of amendment request: The proposed amendment would
revise the Operating License No. NPF-30 for Callaway Plant, Unit 1, in
order to incorporate a change to Technical Specification (TS) 5.5.16,
``Containment Leakage Rate Testing Program,'' which establishes the
program for leakage rate testing of the containment, as required by
Title 10 of Code of Federal Regulations (10 CFR) Section 50.54,
``Conditions of licenses,'' Subsection (o) and 10 CFR 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' Option B, ``Performance Based Requirements,'' as modified
by approved exemptions. Specifically, the TS 5.5.16 would be revised to
reflect a one-time 5-year deferral of the containment Type A integrated
leak rate test (ILRT) from once in 10 years to once in 15 years.
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 will revise Callaway Plant TS 5.5.16,
``Containment Leakage Rate Testing Program,'' to reflect a one-time,
five-year extension for the containment Type A test date to enable
the implementation of a 15-year test interval. While the containment
is designed to contain radioactive material that may be released
from the reactor core following a design basis Loss-of-Coolant
Accident (LOCA), the test interval associated with Type A testing is
part of ensuring the plant's ability to mitigate the consequences of
accidents described in the FSAR [Final Safety Analysis Report] and
does not involve a precursor or initiator of any accident previously
evaluated. Thus, the proposed change to the Type A test interval
cannot increase the probability of an accident previously evaluated
in the FSAR.
Type A testing does provide assurance that the containment will
not exceed allowable leakage rate criteria specified in the TS and
will continue to perform its design function following an accident.
However, per NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' Type A tests identify only a few potential leakage paths
that cannot be identified by Type B and C testing. The current Type
B and C penetration test frequencies for Callaway are established
based on performance, using the requirements of 10 CFR 50, Appendix
J, Option B, and the Type B and C testing requirements will not be
changed as a result of the proposed license amendment. As a result,
with respect to the consequences of an accident, a risk assessment
of the proposed change has concluded that there is an insignificant
increase in total population dose rate and an insignificant increase
in the conditional containment failure probability.
Based on the above, the proposed change does 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 is for a one-time, five-year extension of
the Type A test for Callaway Plant and will not affect the control
parameters governing unit operation or the response of plant
equipment to transient or accident conditions. The proposed change
does not introduce new equipment, modes of system operation, or
failure mechanisms.
Therefore, based on the above, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No
The Callaway Plant containment consists of the concrete
containment building, its steel liner, and the penetrations through
this structure. The structure is designed to contain radioactive
material that may be released from the reactor core following a
design basis LOCA. Additionally, this structure provides shielding
from the fission products that may be present in the containment
atmosphere following accident conditions.
The containment is a prestressed, reinforced concrete,
cylindrical structure with a hemispherical dome and a reinforced
concrete base slab. The inside structure is lined with a carbon
steel liner to ensure a high degree of leak tightness during
operating and accident conditions. A post-tensioning system is used
to prestress the cylindrical shell and dome.
The concrete containment building is required for structural
integrity of the containment under Design Basis Accident (DBA)
conditions. The steel liner and its penetrations establish the
leakage-limiting boundary of the containment. Maintaining
operability of the containment will limit leakage of fission product
radioactivity released from the containment to the environment.
The integrity of the containment penetrations and isolation
valves is verified through Type B and Type C local leak rate tests
(LLRTs) and the overall leak tight integrity of the containment is
verified by an ILRT, as required by 10 CFR 50, Appendix J, ``Primary
Reactor Containment Leakage Testing for Water-Cooled Power
Reactors.''
The existing 10-year interval at Callaway Plant is based on past
performance. Previous Type A tests conducted at Callaway Plant
[[Page 42932]]
indicate that leakage from containment has been less than all 10 CFR
50 Appendix J, Option B, leakage limits.
The proposed change for a one-time extension of the Type A test
does not affect the method for Type A, B, or C testing or the test
acceptance criteria. Type B and C testing will continue to be
performed at the frequency required by Callaway Plant Technical
Specifications. The containment inspections that are performed in
accordance with the requirements of the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section
XI, ``Inservice Inspection,'' and 10 CFR 50.65, ``Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants,'' provide a high degree of a assurance that the containment
will not degrade in a manner that is only detectable by Type A
testing.
In NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' the NRC indicated that a 20-year extension for Type A
testing resulted in an imperceptible increase in risk to the public.
The NUREG-1493 study also concluded that, generically, the design
containment leak rate contributes a very small amount to the
individual risk and that the decrease in Type A testing frequency
would have a minimal affect on this risk. AmerenUE has conducted
risk assessments to determine the impact of a one-time change to the
Callaway Plant Type A test schedule from a baseline value of once in
10 years to once in 15 years for the risk measures of Large Early
Release Frequency (LERF), Total Population Dose, and Conditional
Containment Failure Probability (CCFP). The results of the risk
assessments indicate that the proposed change to the Callaway Plant
Type A test schedule has a minimal impact on public risk.
Based on the above and on previous Type A test results for the
Callaway Plant containment, the current containment surveillance
program, and the results of the AmerenUE risk assessment, there is
no reduction in the effectiveness of the Callaway Plant containment
as a barrier to the release of the post-accident containment
atmosphere to the public or to personnel in the Control Room. Thus,
the proposed changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 4, 2009.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.7.3, ``Main Feedwater Isolation
Valves (MFIVs) and Main Feedwater Regulating Valves (MFRVs), and Main
Feedwater Regulating Valve Bypass Valves (MFRVBVs),'' so that the
limiting condition for operation (LCO) and Applicability more
accurately reflect the conditions for when the LCO should be applicable
and more effectively provide appropriate exceptions to the
Applicability for certain valve configurations. The amendment would
incorporate other minor changes; the title to TS 3.7.3 and the header
for each TS page would be revised, and the exception footnotes in TS
Table 3.3.2-1 of TS 3.3.2, ``ESFAS [Engineered Safety Features
Actuation System] Instrumentation,'' would be revised to improve the
application of existing notes and/or incorporate more appropriate notes
as applicable.
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 do not alter any design or operating
limits, nor do they physically alter safety-related systems, nor do
they affect the way in which safety-related systems perform their
functions. The proposed changes do not change accident initiators or
precursors assumed or postulated in the FSAR [Final Safety Analysis
Report]-described accident analyses, nor do they alter the design
assumptions, conditions, and configuration of the facility or the
manner in which the plant is normally operated and maintained. The
proposed changes do not alter or prevent the ability of structures,
systems, and components (SSCs) from performing their intended
functions to mitigate the consequences of an initiating event within
the assumed acceptance limits. With specific regard to the proposed
TS changes, although the changes involve the exceptions contained in
the Applicability of TS 3.7.3 as well as the notes attached to TS
Table 3.3.2-1 (which are themselves exceptions), the provisions of
the exceptions and notes would continue to be based on the premise
that adequate isolation or isolation capability exists for the main
feedwater lines, i.e., that the required safety function is
performed or capable of being performed as required or assumed for
mitigation of the applicable postulated accidents.
All accident analysis acceptance criteria will therefore
continue to be met with the proposed changes. The proposed changes
will not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. The proposed
changes will not alter any assumptions or change any mitigation
actions in the radiological consequence evaluations in the FSAR. The
applicable radiological dose acceptance criteria will continue to be
met. Overall protection system performance will remain within the
bounds of the previously performed accident analyses.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
There are no proposed design changes, nor are there any changes
in the method by which any safety-related plant structure, system,
or component (SSC) performs its specified safety function. The
proposed changes will not affect the normal method of plant
operation or change any operating parameters. No equipment
performance requirements will be affected. The proposed changes will
not alter any assumptions made in the safety analyses. No new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures will be introduced as a result of this
amendment. There will be no adverse effect or challenges imposed on
any safety-related system as a result of this amendment. The
proposed amendment will not alter the design or performance of the
7300 Process Protection System, Nuclear Instrumentation System, or
Solid State Protection System used in the plant protection systems.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F[Delta]H), loss of coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other margin of safety. The applicable radiological
dose consequence acceptance criteria for design-basis transients and
accidents will continue to be met. The proposed changes do not
eliminate any surveillances or alter the frequency of surveillances
required by the Technical Specifications. None of the acceptance
criteria for any accident analysis will be changed.
[[Page 42933]]
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 4, 2009.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.7.2, ``Main Steam Isolation
Valves (MSIVs),'' to add the main steam isolation valve bypass valves
(MSIVBVs) and main steam low point drain isolation valves (MSLPDIVs) to
the scope of the TS. In addition, the proposed amendment would make
editorial changes to the title and header on each page of TS 3.7.2, and
would incorporate other minor changes to revise exception footnote (i)
in TS Table 3.3.2-1 of TS 3.3.2, ``ESFAS [Engineered Safety Features
Actuation System] Instrumentation,'' to remove the MSIVs from the
footnote such that the footnote only addresses the MSIVBVs and
MSLPDIVs. The MSIVs would be addressed in new exception footnote (k)
added to TS Table 3.3.2-1.
The proposed amendment would add new TS 3.7.19, ``Secondary System
Isolation Valves (SSIVs),'' which would provide limiting conditions for
operation (LCOs) and surveillance requirements for the SSIVs, steam
generator chemical injection isolation valves (SGCIIVs), steam
generator blowdown isolation valves (SGBSIVs), and steam generator
sample line isolation valves (SGBSSIVs). New Function 10, ``Steam
Generator Blowdown System and Sample Line Isolation Valve Actuation,''
would be added to TS Table 3.3.2-1. The SGBSIVs and SGBSSIVs would be
addressed in new exception footnote (t) added to Table 3.3.2-1 for
Function 10.
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 adds requirements to the TS to ensure that
systems and components are maintained consistent with the safety
analysis and licensing basis.
Requirements are incorporated into the TS for secondary system
isolation valves. These changes do not involve any design or
physical changes to the facility, including the SSIVs themselves.
The design and functional performance requirements, operational
characteristics, and reliability of the SSIVs are unchanged. There
is no impact on the design safety function of MSIVs, MSIVBVs,
MSLPDIVs, MFIVs [main feedwater isolation valves], MFRVs [main
feedwater regulating valves] or MFRVBVs [MFRV bypass valves] to
close (either as an accident mitigator or as a potential transient
initiator). Since no failure mode or initiating condition that could
cause an accident (including any plant transient) evaluated per the
FSAR [Final Safety Analysis Report]-described safety analyses is
created or affected, the change cannot involve a significant
increase in the probability of an accident previously evaluated.
With regard to the consequences of an accident and the equipment
required for mitigation of the accident, the proposed changes
involve no design or physical changes to components in the main
steam supply system or feedwater system. There is no impact on the
design safety function of MSIVs, MSIVBVs, MSLPDIVs, MFIVs, MFRVs, or
MFRVBVs or any other equipment required for accident mitigation.
Adequate equipment availability would continue to be required by the
TS. The consequences of applicable, analyzed accidents (such as a
main steam line break [or] feedline break) are not impacted by the
proposed changes.
The changes to TS 3.3.2, TS Table 3.3.2-1, and exception
footnotes associated with Table Function 4 and New Function 10
maintain consistency with the Applicability of revised TS 3.7.2 and
new TS 3.7.19. Maintaining TS 3.3.2 and TS Table 3.3.2-1 consistent
with the Applicability of TS 3.7.2 and TS 3.7.19 is consistent with
the Westinghouse Standard Technical Specifications.
These changes involve no physical changes to the facility and do
not adversely affect the availability of the safety functions
assumed for the MSIVs, MSIVBVs, MSLPDIVs, and SSIVs. Therefore, they
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Based on the above considerations, 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 proposed changes add requirements to the TS that support or
ensure the availability of the safety functions assumed or required
for the MSIVs, MSIVBVs, MSLPDIVs, and SSIVs. The changes do not
involve a physical alteration of the plant (no new or different type
of equipment will be installed) or changes in controlling
parameters. Additional requirements are being imposed, but they are
consistent with the assumptions made in the safety analysis and
licensing basis. The addition of Conditions, Required Actions and
Completion Times to TS for the MSIVBVs, MSLPDIVs, and SSIVs does not
involve a change in the design, configuration, or operational
characteristics of the plant. Further, the proposed changes do not
involve any changes in plant procedures for ensuring that the plant
is operated within analyzed limits. As such, no new failure modes or
mechanisms that could cause a new or different kind of accident from
any previously evaluated are introduced.
Therefore, the proposed changes do 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 addition of Conditions, Required Actions and
Completion Times for SSIVs, MSIVBVs, and MSLPDIVs, as well as the
proposed change to the LCO and Applicability for TS 3.7.2 and the
proposed new TS 3.7.19 (and the corresponding changes to TS 3.3.2,
``ESFAS Instrumentation'') does not alter the manner in which safety
limits or limiting safety system settings are determined. No changes
to instrument/system actuation setpoints are involved. The safety
analysis acceptance criteria are not impacted and the proposed
change will not permit plant operation in a configuration outside
the design basis. The changes are consistent with the safety
analysis and licensing basis for the facility.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: June 1, 2009.
Description of amendment request: The proposed amendment would
revise the Limiting Condition for Operation (LCO) Applicability Note
for Technical Specification (TS) 3.3.9, ``Boron Dilution Mitigation
System (BDMS).''
[[Page 42934]]
The LCO Applicability Note would be revised to more explicitly define
what the term ``during reactor startup'' means in MODES 2 and 3. This
revision to the Applicability Note is proposed to clarify the
situations during which the BDMS signal may be blocked.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no design changes. All design, material, and construction standards
that were applicable prior to this amendment request will be
maintained. There will be no changes to any design or operating
limits.
The proposed change will not adversely affect accident
initiators or precursors [or] adversely alter the design
assumptions, conditions, and configuration of the facility or the
manner in which the plant is operated and maintained. There are no
design or operating changes to the reactor makeup water system
(RMWS), the reactor makeup control system (RMCS), or the chemical
and volume control system (CVCS). There will be no decrease in the
boron concentration of the boric acid tanks. There will be no
changes to the BDMS setpoint or the operation of the BDMS, other
than the limited durations during which flux multiplication signal
blocking would be allowed. Therefore, there will be no changes that
would serve to increase the likelihood of occurrence of an
inadvertent boron dilution event.
The proposed change will not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended functions to mitigate the consequences of an initiating
event within the applicable acceptance limits. Exceptions to
Technical Specification requirements are allowed and, in fact,
rather commonplace when plant operation would otherwise be
restricted in a manner that is not commensurate with the desired
safety objective, especially when those exceptions are of short
duration and are accompanied by compensatory measures.
The proposed change does not physically alter safety-related
systems [or] affect the way in which safety-related systems perform
their functions.
The inadvertent boron dilution analysis acceptance criteria will
continue to be met with the proposed change, with consideration
given to the fact that the current licensing basis analyses do not
assume concurrent rod withdrawal in the MODES 2 and 3 boron dilution
analyses. The licensing basis analyses assume that positive
reactivity insertion is being added by a single method, i.e., boron
dilution. The MODE 2 licensing basis analysis of an inadvertent
boron dilution event in FSAR [Final Safety Analysis Report] Section
15.4.6 assumes that the shutdown banks are fully withdrawn and that
the control banks are withdrawn to the 0% power rod insertion limits
depicted in the COLR [Core Operating Limits Report]. The MODE 2
analysis credits operator action to swap the charging suction source
after an automatic reactor trip, and corresponding rod insertion, on
high source range neutron flux. The MODE 3 licensing basis analysis
credits automatic mitigation by the BDMS with steady state initial
conditions and static initial rod positions (all shutdown and
control banks are fully inserted other than the single most reactive
rod which is assumed to be fully withdrawn) at bounding RCS [reactor
coolant system] T-avg values at either end of MODE 3. Neither the
analysis nor the BDMS design basis assumes that the system protects
against a rod withdrawal event.
The proposed change will not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
The applicable radiological dose criteria will continue to be met.
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.
There are [neither] proposed design changes nor are there any
changes in the method by which any safety-related plant structure,
system, or component (SSC) performs its specified safety function.
The proposed change will not affect the normal method of plant
operation or change any operating parameters. Equipment performance
necessary to fulfill safety analysis missions will be unaffected.
The proposed change will not alter any assumptions required to meet
the safety analysis acceptance criteria.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety-related system as a result of this
amendment.
The proposed amendment will not alter the design or performance
of the 7300 Process Protection System, Nuclear Instrumentation
System, or Solid State Protection System used in the plant
protection systems.
The proposed change does not, therefore, create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F[Delta]H), loss of coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other margin of safety. Mode-specific required
shutdown margins in the COLR will not be changed. The applicable
radiological dose consequence acceptance criteria will continue to
be met.
The proposed change does not eliminate any surveillances or
alter the frequency of surveillances required by the Technical
Specifications. None of the acceptance criteria for any accident
analysis will be changed.
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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: July 10, 2009.
Description of amendment request: The proposed amendment would
delete the Technical Specification (TS) requirements for the
containment hydrogen recombiners and hydrogen monitors. The proposed TS
changes support implementation of the revision to Title 10 of the Code
of Federal Regulations (10 CFR), Section 50.44, ``Standards for
Combustible Gas Control System in Light-Water-Cooled Power Reactors,''
that became effective on October 16, 2003. The proposed changes are
consistent with Revision 1 of the NRC-approved Industry/Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-447, ``Elimination of Hydrogen Recombiners and Change to
Hydrogen and Oxygen Monitors.''
The NRC staff issued a notice of opportunity for public comments on
TSTF-447, Revision 1, published in the Federal Register on August 2,
2002 (67 FR 50374), soliciting comments on a model safety evaluation
(SE) and a model no significant hazards consideration (NSHC)
determination for the elimination of requirements for hydrogen
recombiners, and hydrogen and oxygen monitors from TS. Based on its
evaluation of the public comments
[[Page 42935]]
received, the NRC staff made appropriate changes to the models and
included final versions in a notice of availability published in the
Federal Register on September 25, 2003 (68 FR 55416), regarding the
adoption of TSTF-447, Revision 1, as part of the NRC's consolidated
line item improvement process (CLIIP).
In addition to the changes related to requirements for the hydrogen
recombiners and monitors, this amendment application includes four
unrelated, minor changes to correct typographical errors identified in
Callaway's TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG [Regulatory Guide] 1.97 Category 1 is intended for key variables
that most directly indicate the accomplishment of a safety function
for design-basis accident events. The hydrogen monitors no longer
meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44 the Commission found that Category
3, as defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3 and removal of
the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the SAMGs [severe accident
management guidelines], the emergency plan (EP), the emergency
operating procedures (EOP), and site survey monitoring that support
modification of emergency plan protective action recommendations
(PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this 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 the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the [Three Mile Island],
Unit 2 accident, can be adequately met without reliance on safety-
related hydrogen monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Virginia Electric and Power Company, Docket No. 50-338 North Anna Power
Station, Unit No. 1, Louisa County, Virginia
Date of amendment request: July 23, 2009
Description of amendment request: The proposed change, a one-time
extension to the Completion Time (CT) of Technical Specification 3.8.9
Condition A, will provide an opportunity to fully investigate the
extent of the damaged breaker and its condition to ensure continued bus
reliability for the remainder of the operating cycle.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed change does not alter any plant equipment or
operating practices in such a manner that the probability of an
accident is significantly increased. The proposed change will not
alter assumptions relative to the mitigation of an accident or
transient event. Manual operator actions in the event of an SGTR
have been identified during the one-time extended CT for the 1J1
[Motor Control Center] MCC outage. A risk-informed evaluation of
these operator actions has been performed and the increase in annual
Core Damage and Large Early Release Frequencies associated with the
proposed change in the Technical Specification CT are characterized
as ``small changes'' by Regulatory Guide (RG) 1.174. The Incremental
Conditional Core Damage and Large Early Release Probabilities [ICCDP
and ICLERP] associated with the proposed
[[Page 42936]]
Technical Specification CT meet the acceptance criteria in
Regulatory Guide 1.177.
The ICCDP and ICLERP are 1.01 E-7 per year and 9.86E-9 per year,
respectively. These results are below the RG 1.177 limits of 5E-7
for ICCDP and 5E-8 for ICLERP.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation.
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?
The systems' design and operation are not affected by the
proposed change. The safety analysis acceptance criteria stated in
the Updated Final Safety Analysis Report is not impacted by the
change. Redundancy and diversity of the electrical distribution
system will be maintained with the exception of the MCCs 1J 1-2N and
2S. The proposed change will not allow plant operation in a
configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385
NRC Branch Chief: Undine Shoop.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
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: March 20, 2008, as supplemented by
letters dated May 28, 2008, October 6, 2008, December 17, 2008, and
February 12, 2009.
Brief description of amendment request: The proposed amendments
would revise the McGuire licensing basis by adopting the Alternative
Source Term (AST) radiological analysis methodology as allowed by 10
CFR 50.67, Accident Source Term, for the Loss of Coolant Accident. This
amendment request represents full scope implementation of the AST as
described in Nuclear Regulatory Commission (NRC) Regulatory Guide
1.183, ``Alternative Radiological Source Terms for Evaluating Design
Basis Accidents at Nuclear Power Reactors, Revision 0.''
Date of publication of individual notice in Federal Register:
February 27, 2009 (74 FR 9009).
Expiration date of individual notice: April 28, 2009.
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: June 23, 2008.
Brief description of amendment request: The amendments revise the
Technical Specifications (TSs) for Catawba Nuclear Station, Units 1 and
2. This request modifies the subject TS and Bases by changing the logic
configuration of TS Table 3.3.2-1, ``Engineered Safety Feature
Actuation System Instrumentation'', Function 5.b. (5), ``Turbine Trip
and Feedwater Isolation, Feedwater Isolation, Doghouse Water Level--
High High.'' The existing one-out-of-one (\1/1\) logic per train per
doghouse is being modified to a two-out-of-three (\2/3\) logic per
train per doghouse. The proposed change will improve the overall
reliability of this function and will reduce the potential for spurious
actuations.
Date of publication of individual notice in Federal Register:
February 24, 2009 (74 FR 8276).
Expiration date of individual notice: April 27, 2009.
Duke Energy Carolinas, LLC, et al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of amendment request: November 13, 2008.
Brief description of amendment request: The amendment proposes a
one-cycle revision to the Technical Specifications to incorporate an
interim alternate repair criterion for steam generator tube repair
criteria during the End of Cycle 16 refueling outage and subsequent
cycle 17 operation.
Date of publication of individual notice in Federal Register:
February 24, 2009 (74 FR 8278).
Expiration date of individual notice: April 27, 2009.
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: June 8, 2009.
Brief description of amendment request: The proposed amendment
would revise Technical Specification (TS) 5.5.9.2, ``Unit 1 Model D76
and Unit 2 Model D5 Steam Generator (SG) Program,'' to exclude portions
of the CPSES, Unit 2 Model D5 SG below the top of the SG tubesheet from
periodic SG tube inspections. In addition, the proposed amendment would
revise TS 5.6.9, ``Unit 1 Model D76 and Unit 2 Model D5 Steam Generator
Tube Inspection Report,'' to include reporting requirements specific to
the permanent alternate repair criteria for CPSES, Unit 2. The
amendment request is supported by Westinghouse WCAP-17072-P, ``H*:
Alternate Repair Criteria for the Tube Sheet Expansion Region in Steam
Generators with Hydraulically Expanded Tubes (Model D5),'' May 2009.
Date of publication of individual notice in Federal Register: July
23, 2009 (74 FR 36533).
Expiration date of individual notice: September 21, 2009.
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
[[Page 42937]]
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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: April 23, 2009.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) by removing working hour restrictions
from TS 5.2.2 to support compliance with recent revisions to Title 10
of the Code of Federal Regulations, Part 26, Subpart I. The amendments
are consistent with the guidance contained in Nuclear Regulatory
Commission (NRC) approved Technical Specifications Task Force Traveler
511 (TSTF-511). This TS improvement was made available by the NRC on
December 30, 2008 (73 FR 79923) as part of the consolidated line item
improvement process.
Date of issuance: August 6, 2009.
Effective date: As of the date of issuance to be implemented with
the implementation of the new 10 CFR Part 26, Subpart I requirements.
Amendment Nos.: 292 and 268.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: June 2, 2009 (74 FR
26430).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 6, 2009.
No significant hazards consideration comments received: No.
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 application for amendments: July 14, 2008.
Brief description of amendments: The changes revised Technical
Specifications (TSs) Section 3.7.10, ``Control Room Area Ventilation,''
its associated Bases, and TS Section 5.5 ``Programs and Manuals.'' This
LAR institutes the Control Room Habitability Program.
The changes are consistent with NRC-approved Industry Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-448, Revision 3, ``Control Room Habitability Program.''
The availability of this TS improvement was announced in the Federal
Register on January 17, 2007, as part of the Consolidated Line-Item
Improvement Process (CLIIP). The amendments also authorized a change to
the Catawba Updated Final Safety Analysis Report (UFSAR).
Date of issuance: July 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 250 and 245.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the technical specifications.XXX
Date of initial notice in Federal Register: June 2, 2009 (74 FR
26431).
The Commission's related evaluation, State consultation, and final
no significant hazards consideration determination of the amendments is
contained in a Safety Evaluation dated July 30, 2009.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: January 21, 2009, as supplemented by
letters dated January 23 and June 22, 2009.
Brief description of amendment: The amendment modified the
Technical Specifications (TSs) to adopt U.S. Nuclear Regulatory
Commission (NRC)-approved TS Task Force (TSTF) change travelers TSTF-
163, TSTF-222, TSTF-230, and TSTF-306, and made two minor
administrative corrections.
Date of issuance: August 11, 2009.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 165.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 24, 2009 (74 FR
12392). The supplemental letters dated January 23 and June 22, 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 August 11, 2009.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1 (ANO1), Pope County, Arkansas
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2 (ANO2), Pope County, Arkansas
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAF), Oswego County, New York
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester
County, New York
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant
(PAL), Van Buren County, Michigan
[[Page 42938]]
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station (PIL), Plymouth County, Massachusetts
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana
Parish, Louisiana
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3 (W3), St. Charles Parish, Louisiana
Date of application for amendment: April 27, 2009, as supplemented
July 10, 2009.
Brief description of amendment: The amendments deleted those
portions of the Technical Specifications (TSs) superseded by Title 10
of the Code of Federal Regulations (10 CFR) Part 26, Subpart I,
consistent with U.S. Nuclear Regulatory Commission (NRC)-approved TS
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.''
Date of issuance: August 4, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment Nos.: ANO1--237; ANO2--285; JAF--295; GGNS--183; IP2--
261; IP3--240; PAL--238; PIL--233; RBS--164; and W3--221.
Facility Operating License Nos. DPR-51 (ANO1), NPF-6 (ANO2), DPR-59
(JAF), NPF-29 (GGNS), DPR-26 (IP2), DPR-64 (IP3), DPR-20 (PAL), DPR-35
(PIL), NPF-47 (RBS), and NPF-38 (W3): The amendments revised the
Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 2, 2009 (74 FR
26432). The supplement dated July 10, 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 amendments is contained
in a Safety Evaluation dated August 4, 2009.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 17, 2008, as supplemented by
letters dated January 8, March 18, and June 30, 2009.
Brief description of amendment: The amendment revised the Operating
License and modified Technical Specification (TS) \3/4\.3.1 and Note 2
of TS Table 4.3-1. The changes result in the addition of conservatism
to Core Protection Calculator power indications when calibrations are
required in certain conditions.
Date of issuance: August 10, 2009.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 222.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65695). The supplemental letters dated January 8, March 18, and June
30, 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 August 10, 2009.
No significant hazards consideration comments received: No.
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, 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
Date of application for amendments: June 9, 2008, as supplemented
by letter dated March 30, 2009.
Brief description of amendments: The amendments revise the
Technical Specification (TS) surveillance requirement (SR) frequency in
TS 3.1.3, ``Control Rod OPERABILITY.'' The amendments also clarify the
requirement to fully insert all insertable control rods for the
limiting condition for operation in TS 3.3.1.2, Required Action E.2,
``Source Range Monitoring Instrumentation'' (Clinton Power Station
only). Finally, the amendments revise Example 1.4-3 in Section 1.4,
``Frequency,'' to clarify the applicability of the 1.25 surveillance
test interval extension.
Date of issuance: August 11, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 188, 232/225, 193/180, 272/276, 244/239.
Facility Operating License Nos. NPF-62, DPR-19, DPR-25, NPF-11,
NPF-18, DPR-44, DPR-56, DPR-29, DPR-30: The amendments revised the
Technical Specifications/Licenses.
Date of initial notice in Federal Register: August 12, 2009 (73 FR
46928) The March 30, 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 August 11, 2009.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Unit Nos. 1 and 2 (CPSES),
Somervell County, Texas
Date of amendment request: April 1, 2009, as supplemented by letter
dated July 9, 2009.
Brief description of amendments: The amendments deleted Technical
Specification (TS) 5.2.2.d, in TS 5.2.2, ``Unit Staff,'' regarding the
requirement to develop and implement administrative procedures to limit
the working hours of personnel who perform safety-related functions. In
addition, paragraphs e and f of TS 5.2.2 were renumbered to d and e and
in TS 5.2.2.b the reference to 5.2.2.f was revised to 5.2.2.e to
reflect the removal of paragraph d of TS 5.2.2. The change is
consistent with U.S. Nuclear Regulatory Commission (NRC)-approved
Revision 0 to TS Task Force (TSTF) Improved Technical Specification
change traveler, TSTF-511, ``Eliminate Working Hour Restrictions from
TS 5.2.2 to Support Compliance with 10 CFR Part 26.'' The availability
of this TS improvement was announced in the Federal Register on
December 30, 2008 (73 FR 79923), as part of the consolidated line item
improvement process.
Date of issuance: August 7, 2009.
[[Page 42939]]
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-148; Unit 2-148.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 19, 2009 (74 FR
23445). The supplemental letter dated July 9, 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 amendments is contained
in a Safety Evaluation dated August 7, 2009.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP 1 and 2),
Oswego County, New York
Date of application for amendment: February 11, 2009.
Brief description of amendments: The amendments delete those
portions of the Technical Specifications (TSs) superseded by Title 10
of the Code of Federal Regulations (10 CFR), Part 26, Subpart I. This
change is consistent with Nuclear Regulatory Commission (NRC) approved
Technical Specification Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler TSTF-511, Revision 0, ``Eliminate
Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10
CFR Part 26.'' These changes were described in a Notice of Availability
for Consolidated Line Item Improvement Process TSTF-511 published in
the Federal Register on December 30, 2008 (73 FR 79923).
Date of issuance: July 27, 2009.
Effective date: As of the date of issuance to be implemented by
October 1, 2009.
Amendment Nos.: 203 and 131.
Renewed Facility Operating License Nos. DPR-063 and NPF-069: The
amendments revise the License and TSs.
Date of initial notice in Federal Register: April 21, 2009 (73 FR
18255).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 2009.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: September 2, 2008.
Brief description of amendments: The amendment approved the
licensee's request to incorporate a revision in the Updated Final
Safety Analysis Report (UFSAR) Section 13.7.2.3, ``PRA Risk
Categorization,'' to add a separate set of criteria for assessing the
risk significance of the risk achievement worth values of common cause
failures as part of the probabilistic risk assessment analysis of the
risk importance of components.
Date of issuance: August 12, 2009
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1-191; Unit 2-179.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses, and Updated Final Safety
Analysis Report.
Date of initial notice in Federal Register: December 2, 2008 (73 FR
73354).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 12, 2009.
No significant hazards consideration comments received: No.
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 amendments: February 6, 2009.
Brief description of amendments: The proposed amendments deleted
applicable portions of the Technical Specifications (TSs) superseded by
Part 26, Subpart I of Title 10 of the Code of Federal Regulations (10
CFR). This change is consistent with Nuclear Regulatory Commission
(NRC)-approved Revision 0 to Technical Specification Task Force (TSTF)
Improved Standard Technical Specification Change Traveler, TSTF-511,
``Eliminate Working Hour Restrictions from TS 5.2-2 to Support
Compliance with 10 CFR Part 26.''
Date of issuance: July 29, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment Nos.: 256 and 237.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the license and the technical specifications.
Date of initial notice in Federal Register: March 24, 2009 (74 FR
12396).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 2009.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: August 14, 2008.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' to extend the Surveillance Frequency on
selected ESFAS slave relays from 92 days to 18 months.
Date of issuance: July 30, 2009.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 183.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 7, 2008 (73 FR
58379).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 30, 2009.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: August 14, 2008, as supplemented by
letter dated April 10, 2009.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' TS 3.7.2, ``Main Steam Isolation Valves
(MSIVs),'' and added new TS 3.7.19, ``Secondary System Isolation Valves
(SSIVs).'' TS 3.7.2 has been revised to add MSIV bypass valves to the
scope of TS 3.7.2. TS Table 3.3.2-1 has been revised to reflect the
addition of the MSIV bypass valves to TS 3.7.2 and the associated
applicability to be consistent with Westinghouse Standard Technical
Specifications (NUREG-1431, Revision 3.0). TS 3.7.19 has been added to
include a limiting condition for operation, conditions/required
actions, and surveillance requirements for the steam generator blowdown
isolation valves and steam generator blowdown sample isolation valves.
[[Page 42940]]
Date of issuance: July 31, 2009.
Effective date: As of the date of issuance and shall be implemented
prior to startup from Refueling Outage 17.
Amendment No.: 184.
Renewed Facility Operating License No.: NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 7, 2008 (73 FR
58679). The supplemental letter dated April 10, 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 July 31, 2009.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 6, 2009, as supplemented by letter
dated July 14, 2009.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.2.2, ``Unit Staff,'' to eliminate working hour
restrictions (TS 5.2.2.d) to support compliance with Title 10 of the
Code of Federal Regulations (10 CFR) Part 26. In addition, paragraphs e
and f of TS 5.2.2 were renumbered to d and e to reflect the removal of
paragraph d of TS 5.2.2, and a reference in 5.2.2b was updated to
reflect the renumbering of 5.2.2f. to 5.2.2e. The request is consistent
with the guidance contained in U.S. Nuclear Regulatory Commission
(NRC)-approved TS 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.''
Date of issuance: August 7, 2009.
Effective date: As of its date of issuance and shall be implemented
by October 1, 2009.
Amendment No.: 185.
Renewed Facility Operating License No.: NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 21, 2009 (74 FR
18258). The supplemental letter dated July 14, 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 August 7, 2009.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 14th day of August 2009.
For the Nuclear Regulatory Commission.
Allen G. Howe,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E9-20403 Filed 8-24-09; 8:45 am]
BILLING CODE 7590-01-P