[Federal Register Volume 74, Number 16 (Tuesday, January 27, 2009)]
[Notices]
[Pages 4764-4781]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-1568]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0016]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 31, 2008 to January 13, 2009. The
last biweekly notice was published on January 13, 2009 (74 FR 1712).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example, in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
[[Page 4765]]
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Copies of written comments
received may be examined at the Commission's Public Document Room
(PDR), located at One White Flint North, Public File Area O1F21, 11555
Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve all adjudicatory documents
over the Internet or in some cases to mail copies on electronic storage
media. Participants may not submit paper copies of their filings unless
they seek a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer \TM\ to
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available
at http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
Information about applying for a digital ID certificate is available on
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice
[[Page 4766]]
confirming receipt of the document. The EIE system also distributes an
e-mail notice that provides access to the document to the NRC Office of
the General Counsel and any others who have advised the Office of the
Secretary that they wish to participate in the proceeding, so that the
filer need not serve the documents on those participants separately.
Therefore, applicants and other participants (or their counsel or
representative) must apply for and receive a digital ID certificate
before a hearing request/petition to intervene is filed so that they
can obtain access to the document via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday. The help electronic filing Help Desk can be
contacted by telephone at 1-866-672-7640 or by e-mail at
[email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) First class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: July 2, 2008.
Description of amendment request: The amendments would revise
Technical Specification (TS) 4.2.2, ``Control Element Assemblies,'' to
support replacement of the full strength control element assemblies
(CEAs) with a new design beginning with the 14th refueling outage
(U3R14) for Palo Verde Nuclear Generating Station (PVNGS), Unit 3 in
the spring of 2009. Additionally, Arizona Public Service Company (APS)
will be updating the TS by removing the registered trademark
``Inconel'' while retaining the generic terminology ``Alloy 625'' and
deleting the references to part-length CEAs in TS 4.2.2.
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?
Replacement of full-strength compression sleeve control element
assemblies with full-strength silver (Ag)-indium (In)-Cadmium (Cd)
control element assemblies.
Response: No.
The proposed change involves a new design for the full-strength
Control Element Assemblies (CEA) that replaces a portion of B4C
pellets (including the compression sleeve) in the tips of the CEA
fingers with hollow silver-indium-cadmium slugs.
The following events are related to inadvertent movement of the
CEAs; however, they are not initiated by the CEAs.
Uncontrolled Control Element Assembly Withdrawal from a
Subcritical or Low (Hot Zero) Power Condition.
Uncontrolled Control Element Assembly Withdrawal at
Power.
Single Full-Strength Control Element Assembly Drop.
Control Element Assembly Ejection.
These previously analyzed accidents are initiated by the failure
of plant structures, systems, or components (SSC) other than the CEA
itself. The proposed change to the CEA design does not have a
detrimental impact on the integrity of any plant SSC that initiates
an analyzed event. Additionally, the CEAs mitigate other events. In
these events, the chrome plating on the portion of the clad exterior
and the added weight has been conservatively accounted for in the
SCRAM [safety control rod axe man] calculation. The change does not
adversely affect the protective and mitigative capabilities of the
plant, nor does the change affect the initiation or probability of
occurrence of any accident. The SSCs will continue to perform their
intended safety functions.
The proposed change in CEA design has resulted in a slight (less
than 1%) reduction of total reactivity.
Computer modeling events which exhibit sensitivity to time
dependent rod worth (sheared shaft/seized rotor, loss of flow from
SAFDL [specified acceptable fuel design limits] and total loss of
reactor coolant flow) demonstrate that all acceptance criteria
continued to be met.
Therefore this change will not significantly increase the
probability or consequences of any accident previously evaluated.
The removal of the registered trademark name ``Inconel''.
Response: No.
This change is considered editorial. Inconel is a registered
trademark of Special Metals Corporation, while Alloy 625 is a
generic alloy designation from the Unified Numbering System.
Retaining the already referenced term ``Alloy 625'' does not involve
a significant increase in the probability or consequences of an
accident previously evaluated, as the material properties and
application of Alloy 625 have not changed.
Deletion of the references to part-length control element
assemblies.
Response: No.
This change is considered editorial. The removal of this
information does not involve a significant increase in the
probability or consequences of an accident previously evaluated as
the part-length CEAs were
[[Page 4767]]
replaced in accordance with License Amendment 152, dated March 23,
2004 (Agency Document Access and Management System (ADAMS) Accession
No. ML040860573) and the information is no longer applicable.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Replacement of full-strength compression sleeve control element
assemblies with full-strength silver(Ag)-indium(In)-Cadmium(Cd)
control element assemblies.
Response: No.
There are three differences in the replacement CEAs as compared
to the current CEAs.
First, there is a very slight change in the outside diameter of
a portion of the cladding on the replacement CEAs due to chrome
plating on the lower portion of cladding. Analysis demonstrates that
this change will not cause interference between the CEA cladding and
the guide tube inside diameter in the buffer region. Secondly, there
is a slight increase in weight with the Ag-In-Cd CEAs. However, this
difference has been analyzed with respect to the performance
capability of the CEDMs [Control Element Drive Mechanisms] and found
to be within design capabilities and design analyses. Finally, the
upper edges of the spider bosses have been chamfered to prevent
damage to the self-latching mechanisms that can occur if the CEA
hangs up when lifting through the upper guide structure cut outs.
This change is for ease of maintenance and has no impact on
operation of the CEAs.
Therefore, the Ag-In-Cd CEAs are identical to the compression
sleeve CEAs in terms of form, fit and function and the proposed
change will not introduce any new failure mechanisms, malfunctions,
or accident initiators not already considered in the design and
licensing bases. The possibility of a new or different malfunction
of safety-related equipment is not created. No new accident
scenarios, transient precursors, or limiting single failures are
introduced as a result of these changes. There will be no adverse
effects or challenges imposed on any safety-related system as a
result of these changes. Therefore, the possibility of a new or
different accident from any accident previously evaluated is not
created as a result of any dimensional change.
The removal of the registered trademark name ``Inconel''.
Response: No.
This change is considered editorial. Inconel is a registered
trademark of Special Metals Corporation, while Alloy 625 is a
generic alloy designation from the Unified Numbering System.
Retaining the already referenced term ``Alloy 625'' does not create
the possibility of a new or different kind of accident from any
accident previously evaluated, as the material properties and
application of Alloy 625 have not changed.
Deletion of the references to part-length control element
assemblies.
Response: No.
This change is considered editorial. The removal of this
information does not create the possibility of a new or different
kind of accident from any accident previously evaluated as the part-
length CEAs were replaced in accordance with License Amendment 152,
dated March 23, 2004 (Agency Document Access and Management System
(ADAMS) Accession No. ML040860573) and the information is no longer
applicable.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Replacement of full-strength compression sleeve control element
assemblies with full-strength silver(Ac)-indium(In)-Cadmium(Cd)
control element assemblies.
Response: No.
Reactor core safety limits are established in the PVNGS
Technical Specifications to prevent overheating of the fuel and
cladding that would result in the release of fission products to the
reactor coolant during steady state operation, normal operational
transients, and anticipated operational occurrences. The margin to
these safety limits is not affected by the CEA design changes under
consideration.
Overheating of the fuel is prevented by maintaining steady
state, peak linear heat rate (LHR) below the level at which fuel
centerline melting occurs. If the local LHR is high enough to cause
the fuel centerline temperature to reach the melting point of the
fuel, expansion of the pellet caused by centerline melting may cause
the pellet to stress the cladding to the point of failure, allowing
an uncontrolled release of activity to the reactor coolant.
Compliance with the DNBR [departure from nucleate boiling ratio]
and fuel centerline melt specified acceptable fuel design limits
(SAFDLs) is assured through the CEA insertion limits and alignment
technical specifications, and through the power distribution limit
technical specifications.
There is no change to the operation of the full-strength CEAs
due to the change from compression sleeve CEAs to Ag-In-Cd CEAs.
Since the Ag-In-Cd CEAs may be used to control power distribution
similar to the compression sleeve CEAs, power distributions will
still be controlled and maintained within the limits necessary to
assure SAFDLs are met.
The proposed change in CEA design has resulted in a slight (less
than 1%) reduction in total reactivity.
Computer modeling results of events which exhibit sensitivity to
time dependent rod worth (sheared shaft/seized rotor, loss of flow
from SAFDL and total loss of reactor coolant flow) demonstrate that
all acceptance criteria continued to be met.
Therefore, since SAFDLs continue to be met, the change from
compression sleeve CEAs to Ag-In-Cd CEAs does not involve a
significant reduction in a margin of safety.
The removal of the registered trademark name ``Inconel''.
Response: No.
The removal of the registered trademark name ``Inconel'' [ ] is
considered editorial. Inconel is a registered trademark of Special
Metals Corporation, while Alloy 625 is a generic alloy designation
from the Unified Numbering System. Retaining the already referenced
term ``Alloy 625'' does not involve a significant reduction in the
margin of safety as the material properties and application of Alloy
625 have not changed.
Deletion of the references to part-length control element
assemblies.
Response: No.
This change is considered editorial. The removal of this
information does not involve a significant reduction in the margin
of safety as the part-length CEAs were replaced in accordance with
Amendment 152, dated March 23, 2004 (Agency Document Access and
Management System (ADAMS) Accession No. ML040860573) and the
information is no longer applicable.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: October 6, 2008.
Description of amendments request: The proposed change would remove
work hour controls and/or references to the NRC Generic Letter 82-12
from the administrative control sections of the technical
specifications. On April 17, 2007, the NRC approved a final rule that
amended 10 CFR Part 26 and, among other changes, established
requirements for managing worker fatigue at operating nuclear power
plants. Subpart I, ``Managing Fatigue,'' specifically addresses
managing worker fatigue by designating individual break requirements,
work hour limits, and annual reporting requirements. Subpart I was
published in the Federal Register on March 31, 2008 (73 FR 16966), with
a required implementation period of 18 months. Compliance is,
therefore, required by October 1, 2009. In order to support compliance
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove
these work hour controls from Technical Specification 5.2.2.e at the
Brunswick Steam Electric Plant, Units 1 and 2.
Basis for proposed no significant hazards consideration
determination Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 4768]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes remove TS [technical specification]
controls on working hours for personnel who perform safety related
functions. The TS controls are superseded by the worker fatigue
requirements in 10 CFR Part 26. Removal of the TS requirements will
be performed concurrently with the implementation of the 10 CFR Part
26, Subpart I requirements. The proposed changes do not impact the
physical configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. The proposed changes do
not impact the initiators or assumptions of analyzed events, nor do
they impact the mitigation of accidents or transient events.
Therefore, it is concluded that these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26.
Work hours will continue to be controlled in accordance with NRC
requirements. The new rule allows for deviations from controls to
mitigate or prevent a condition adverse to safety or as necessary to
maintain the security of the facility. This ensures that the new
rule will not restrict work hours and thereby create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter plant configuration, require
that new plant equipment be installed, alter assumptions made about
accidents previously evaluated, add any initiators, or effect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26. The
proposed changes do not involve any physical changes to plant or the
manner in which plant systems are operated, maintained, modified,
tested, or inspected. The proposed changes do not alter the manner
in which safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
changes will not result in plant operation in a configuration
outside the design basis. The proposed changes will not adversely
affect systems that respond to safely shutdown the plant and to
maintain the plant in a safe shutdown condition.
Removal of plant-specific TS administrative requirements will
not reduce a margin of safety because the requirements in 10 CFR
Part 26 are adequate to ensure that worker fatigue is managed.
Therefore, it is concluded that these changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: October 6, 2008.
Description of amendments request: The proposed change would remove
work hour controls and/or references to the NRC Generic Letter 82-12
from the administrative control sections of the technical
specifications. On April 17, 2007, the NRC approved a final rule that
amended 10 CFR Part 26 and, among other changes, established
requirements for managing worker fatigue at operating nuclear power
plants. Subpart I, ``Managing Fatigue,'' specifically addresses
managing worker fatigue by designating individual break requirements,
work hour limits, and annual reporting requirements. Subpart I was
published in the Federal Register on March 31, 2008 (73 FR 16966), with
a required implementation period of 18 months. Compliance is,
therefore, required by October 1, 2009. In order to support compliance
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove
these work hour controls from Technical Specification 5.2.2.e at the H.
B. Robinson Steam Electric Plant, Unit 2.
Basis for proposed no significant hazards consideration
determination Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes remove TS [technical specification]
controls on working hours for personnel who perform safety related
functions. The TS controls are superseded by the worker fatigue
requirements in 10 CFR Part 26. Removal of the TS requirements will
be performed concurrently with the implementation of the 10 CFR Part
26, Subpart I requirements. The proposed changes do not impact the
physical configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. The proposed changes do
not impact the initiators or assumptions of analyzed events, nor do
they impact the mitigation of accidents or transient events.
Therefore, it is concluded that these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26.
Work hours will continue to be controlled in accordance with NRC
requirements. The new rule allows for deviations from controls to
mitigate or prevent a condition adverse to safety or as necessary to
maintain the security of the facility. This ensures that the new
rule will not restrict work hours and thereby create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter plant configuration, require
that new plant equipment be installed, alter assumptions made about
accidents previously evaluated, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26. The
proposed changes do not involve any physical changes to the plant or
the manner in which plant systems are operated,
[[Page 4769]]
maintained, modified, tested, or inspected. The proposed changes do
not alter the manner in which safety limits, limiting safety system
settings or limiting conditions for operation are determined. The
safety analysis acceptance criteria are not affected by this change.
The proposed changes will not result in plant operation in a
configuration outside the design basis. The proposed changes will
not adversely affect systems that respond to safely shut down the
plant and to maintain the plant in a safe shutdown condition.
Removal of plant-specific TS administrative requirements will
not reduce a margin of safety because the requirements in 10 CFR
Part 26 are adequate to ensure that worker fatigue is managed.
Therefore, it is concluded that these changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: October 6, 2008.
Description of amendment request: The proposed change would remove
work hour controls and/or references to the NRC Generic Letter 82-12
from the administrative control sections of the technical
specifications. On April 17, 2007, the NRC approved a final rule that
amended 10 CFR Part 26 and, among other changes, established
requirements for managing worker fatigue at operating nuclear power
plants. Subpart I, ``Managing Fatigue,'' specifically addresses
managing worker fatigue by designating individual break requirements,
work hour limits, and annual reporting requirements. Subpart I was
published in the Federal Register on March 31, 2008 (73 FR 16966), with
a required implementation period of 18 months. Compliance is,
therefore, required by October 1, 2009. In order to support compliance
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove
these work hour controls from Technical Specification 6.2.2.f at the
Shearon Harris Nuclear Power Plant, Unit 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes remove TS [technical specification]
controls on working hours for personnel who perform safety related
functions. The TS controls are superseded by the worker fatigue
requirements in 10 CFR Part 26. Removal of the TS requirements will
be performed concurrently with the implementation of the 10 CFR Part
26, Subpart I requirements. The proposed changes do not impact the
physical configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. The proposed changes do
not impact the initiators or assumptions of analyzed events, nor do
they impact the mitigation of accidents or transient events.
Therefore, it is concluded that these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26.
Work hours will continue to be controlled in accordance with NRC
requirements. The new rule allows for deviations from controls to
mitigate or prevent a condition adverse to safety or as necessary to
maintain the security of the facility. This ensures that the new
rule will not restrict work hours and thereby create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter plant configuration, require
that new plant equipment be installed, alter assumptions made about
accidents previously evaluated, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26. The
proposed changes do not involve any physical changes to the plant or
the manner in which plant systems are operated, maintained,
modified, tested, or inspected. The proposed changes do not alter
the manner in which safety limits, limiting safety system settings
or limiting conditions for operation are determined. The safety
analysis acceptance criteria are not affected by this change. The
proposed changes will not result in plant operation in a
configuration outside the design basis. The proposed changes will
not adversely affect systems that respond to safely shut down the
plant and to maintain the plant in a safe shutdown condition.
Removal of plant-specific TS administrative requirements will
not reduce a margin of safety because the requirements in 10 CFR
Part 26 are adequate to ensure that worker fatigue is managed.
Therefore, it is concluded that these changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: November 13, 2008.
Description of amendment request: The proposed change will modify
Technical Specification (TS) 3.3.1.1, ``Reactor Protective
Instrumentation.'' Specifically, Table 4.3-1 and the associated Notes 7
and 8 will be revised to clarify and streamline the reactor coolant
system (RCS) flow verification requirements associated with the
departure from nucleate boiling ratio (DNBR) reactor trip signal.
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 CPC [Core Protection Calculator] reactor protective function
is not considered an accident initiator. The primary function is to
initiate an automatic reactor trip signal when specific plant
conditions are reached, thereby limiting the consequences of an
accident. The proposed change acts to eliminate unnecessary
conservatisms and accordingly increase operational margin by
eliminating the requirement to use
[[Page 4770]]
calorimetric flow measurement in the CPC flow verification. This
method of verification will normally only be used in the future
during periods when the COLSS [Core Operating Limits Supervisory
System] RCP [Reactor Coolant Pump] [Delta] p flow measurement is
unavailable. Regardless of the method of verification used, the CPC
will continue to be verified to have an indicated RCS flow equal to
or conservative relative to the measured RCS flow on a once per 12-
hour basis. In so doing, the CPC will continue to act to generate a
reactor trip on low DNBR as originally designed in order to ensure
the DNBR reactor core Safety Limit is not exceeded.
The relocation of measurement uncertainty references to the TS
Bases does not reduce the requirements to account for uncertainties
in any Limiting Safety System Setting (LSSS) designed to protect
reactor core Safety Limits. The necessary uncertainties will
continue to be applied as required and will be controlled in
accordance with TS 6.5.14, Technical Specification Bases Control
Program, and station procedures.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in any physical plant
modifications or changes in the way the plant is operated. In
addition, the CPCs are unrelated to any type of accident initiator
previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change increases operating margin when the COLSS
RCP [Delta]p flow measurement is available for use while unaffecting
the CPC ability to initiate an automatic reactor trip on low DNBR
prior to the DNBR reactor core safety limit being exceeded.
Relocating the references to measurement uncertainties to the TS
Bases likewise has no impact on the CPC design function and the
uncertainties will continue to be applied as required and controlled
in accordance with TS 6.5.14, Technical Specification Bases Control
Program, and station procedures.
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.
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: December 8, 2008
Description of amendment request: The proposed amendment adds a
license condition to allow a one-time extension of surveillance
requirements involving the 18-month channel calibration and logic
system functional tests for one channel of the reactor water level
instrumentation system. The extension is to account for the effects of
rescheduling the next refueling outage from early to late 2009.
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 requested action is a one-time extension to the performance
interval of certain TS [Technical Specification] surveillance
requirements. The performance of the surveillances, or the failure
to perform the surveillances, is not a precursor to an accident.
Performing the surveillances or failing to perform the surveillances
does not affect the probability of an accident. Therefore, the
proposed delay in performance of the surveillance requirements in
this amendment request does not increase the probability of an
accident previously evaluated.
A delay in performing the surveillances does not result in a
system being unable to perform its required function. Additionally,
the defense in depth of the system design provides additional
confidence that the safety function is maintained. In the case of
this one-time extension request, the relatively short period of
additional time that the systems and components will be in service
before the next performance of the surveillance will not affect the
ability of those systems to operate as designed. Therefore, the
systems required to mitigate accidents will remain capable of
performing their required function. No new failure modes have been
introduced because of this action and the consequences remain
consistent with previously evaluated accidents. Therefore, the
proposed delay in performance of the surveillance requirement in
this amendment request does not involve a significant increase in
the consequences of 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of
any system, structure, or component (SSC), or a change in the way
any SSC is operated. The surveillance intervals of the level
instrumentation are currently evaluated for 30 months, which bounds
the requested interval extension. The proposed amendment does not
involve operation of any SSCs in a manner or configuration different
from those previously recognized or evaluated. No new failure
mechanisms will be introduced by the one-time surveillance extension
being requested.
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 amendment is a one-time extension of the
performance-interval of certain TS surveillance requirements.
Extending the surveillance requirements does not involve a
modification of any TS Limiting Conditions for Operation. Extending
the surveillance frequency does not involve a change to any limit on
accident consequences specified in the license or regulations.
Extending the surveillance frequency does not involve a change to
how accidents are mitigated or a significant increase in the
consequences of an accident. Extending the surveillance frequency
does not involve a change in a methodology used to evaluate
consequences of an accident. Extending the surveillance frequency
does not involve a change in any operating procedure or process. The
surveillance intervals of the level instrumentation are currently
evaluated for 30 months which bounds the requested interval
extension. The components involved in this request have exhibited
reliable operation based on the results of the most recent
performances of their 18-month surveillance requirements and the
associated functional surveillances.
Based on the limited additional period of time that the systems
and components will be in service before the surveillance is next
performed, as well as the operating experience that these
surveillances are typically successful when performed, it is
reasonable to conclude that the margin of safety associated with the
surveillance requirement will not be affected by the requested
extension.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 4771]]
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.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1
and 2, Ogle County, Illinois.
Date of amendment request: December 4, 2008.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16,
``RCS Specific Activity,'' and Surveillance Requirements 3.4.16.1 and
3.4.16.3. The proposed changes would replace the current TS 3.4.16
limit on reactor coolant system (RCS) gross specific activity with a
new limit on RCS noble gas specific activity. The noble gas specific
activity limit would be based on a new dose equivalent Xe-133
definition that would replace the current E Bar average disintegration
energy definition. In addition, the current dose equivalent I-131
definition would be reformatted. The availability of this TS revision
was announced in the Federal Register on March 15, 2007 (72 FR 12217)
as part of the consolidated line item improvement process. The licensee
affirmed the applicability of the model no significant hazards
consideration determination in its application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration adopted by the licensee is
presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
Reactor coolant specific activity is not an initiator for any
accident previously evaluated. The Completion Time when primary coolant
gross activity is not within limit is not an initiator for any accident
previously evaluated. The current variable limit on primary coolant
iodine concentration is not an initiator to any accident previously
evaluated. As a result, the proposed change does not significantly
increase the probability of an accident. The proposed change will limit
primary coolant noble gases to concentrations consistent with the
accident analyses. The proposed change to the Completion Time has no
impact on the consequences of any design basis accident since the
consequences of an accident during the extended Completion Time are the
same as the consequences of an accident during the Completion Time. As
a result, the consequences of any accident previously evaluated are not
significantly increased.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident from any Accident Previously
Evaluated.
The proposed change in specific activity limits does not alter any
physical part of the plant nor does it affect any plant operating
parameter. The change does not create the potential for a new or
different kind of accident from any previously calculated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change revises the limits on noble gas radioactivity
in the primary coolant. The proposed change is consistent with the
assumptions in the safety analyses and will ensure the monitored values
protect the initial assumptions in the safety analyses.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
analysis adopted by the licensee and, based on this review, it appears
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendments involve no
significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: September 2, 2008.
Description of amendment request: The proposed amendments would
relocate Surveillance Requirements (SR) 3.8.3.6 from the technical
specifications (TSs) to a licensee-controlled document. SR 3.8.3.6
requires Emergency Diesel Generator fuel oil storage tanks to be
drained, sediment removed, and cleaned on a 10-year interval. The
change is consistent with the current revision (i.e., Rev. 3) of the
Improved Standard Technical Specifications (ISTS), NUREG 1434,
``Standard Technical Specifications General Electric Plants, BWR/6.''
The SR was removed from the ISTS under Technical Specification Task
Force (TSTF) Traveler No. 2, ``Relocate the 10-Year Sediment Cleaning
of the Fuel Oil Storage Tank to Licensee Control,'' approved by the
Nuclear Regulatory Commission on July 16, 1998.
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 FOSTs [fuel oil storage tanks] provide the storage for the
DG [diesel generator] fuel oil, assuring an adequate volume is
available for each DG to operate for seven days in the event of a
loss of offsite power concurrent with a loss of coolant accident.
The relocation of the SR to drain and clean the FOSTs to a licensee-
controlled document will not impact any of the previously analyzed
accidents. Sediment in the tank, or failure to perform this SR, does
not necessarily result in an inoperable storage tank. Fuel oil
quantity and quality are assured by other TS SRs that remain
unchanged. These SRs help ensure tank sediment is minimized and
ensure that any degradation of the tank wall surface that results in
fuel oil volume reduction is detected and corrected in a timely
manner. Future changes to the licensee-controlled document will be
evaluated pursuant to the requirements of 10 CFR 50.59, ``Changes,
tests, and experiments,'' to ensure that such changes do not result
in more than a minimal increase in the probability or consequences
of an accident previously evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration or the manner in which the plant is
operated and maintained. The proposed change does not adversely
affect the ability of structures, systems or components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits.
The proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed change does not increase the types and amounts
of radiological effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
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 TS change does not involve the addition or
modification of any plant equipment. Also, the proposed change will
[[Page 4772]]
not alter the design configuration, or method of operation of plant
equipment beyond its normal functional capabilities. The
requirements retained in the TS continue to require testing of the
diesel fuel oil to ensure the proper functioning of the DGs. The
proposed TS change does not create any new credible failure
mechanisms, malfunctions or accident initiators.
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 proposed change does not alter or exceed a design basis or
safety limit. The requirements retained in the TS continue to
require testing of the diesel fuel oil to ensure the DGs are able to
perform their intended function.
Therefore, the proposed changes 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: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: September 24, 2008.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TSs) to allow the BVPS-2 containment
spray additive sodium hydroxide (NaOH) to be replaced by sodium
tetraborate (NaTB).
Basis for proposed no significant hazards consideration
determination: As required by10 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.
Use of NaTB in lieu of NaOH would not involve a significant
increase in probability of a previously evaluated accident because
the containment spray additive is not an initiator of any analyzed
accident. The NaTB would be stored and delivered by a passive method
that does not have potential to affect plant operations. Any
existing NaOH delivery system equipment which remains in place but
is removed from service would meet existing seismic, electrical and
containment isolation requirements. Therefore the change in
additive, including removal of NaOH equipment from service, would
not result in any failure modes that could initiate an accident.
The spray additive is used to mitigate the consequences of a
LOCA [loss-of-coolant accident]. Use of NaTB as an additive in lieu
of NaOH would not involve a significant increase in the consequences
of a previously evaluated accident because the amount of NaTB
specified in the proposed TS would achieve a pH of 7 or greater,
consistent with the current licensing basis. This pH is sufficient
to achieve long-term retention of iodine by the containment sump
fluid for the purpose of reducing accident related radiation dose
following a LOCA.
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.
Regarding the proposed use of NaTB in lieu of NaOH, the NaTB
would be stored and delivered by a passive method that does not have
potential to affect plant operations. Any existing NaOH delivery
system equipment remaining in place but which is removed from
service would meet existing seismic, electrical and containment
isolation requirements. Hydrogen generation would not be
significantly impacted by the change. Therefore, no new failure
mechanisms, malfunctions, or accident initiators would be introduced
by the proposed change and it would 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.
Since the quantity of NaTB specified in the amended TS would
reduce the potential for undesirable chemical effects while
achieving radiation dose reductions, corrosion control and hydrogen
generation effects that are comparable to NaOH, the proposed change
does not involve a significant reduction in a margin of safety. The
primary function of an additive is to reduce loss of coolant
accident consequences by controlling the amount of iodine fission
products released to containment atmosphere from reactor coolant
accumulating in the sump during a LOCA. Because the amended
technical specifications would achieve a pH of 7 or greater using
NaTB, dose related safety margins would not be significantly
reduced. Use of NaTB reduces the potential for undesirable chemical
effects that could interfere with recirculation flow through the
sump strainers. Any existing NaOH delivery system equipment which
remains in place but is removed from service would meet existing
seismic, electrical and containment isolation requirements and would
not interfere with operation of the existing containment or
containment spray system.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: Mark G. Kowal.
FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit No. 1 (PNPP), Lake County, Ohio
Date of amendment request: November 18, 2008
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 5.5.6 to incorporate Technical
Specification Task Force (TSTF) Travelers TSTF-479, ``Changes to
Reflect Revision of 10 CFR 50.55a,'' and TSTF 497, ``Limit Inservice
Testing Program SR [Surveillance Requirement] 3.0.2 Application to
Frequencies of 2 Years or Less.''
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 amendment revises TS 5.5.6, ``Inservice Testing
Program,'' for consistency with 10 CFR 50.55a(f)(4) requirements
regarding inservice testing of pumps and valves. The proposed
amendment incorporates revisions to the ASME Code that result in a
net improvement in the measures for testing pumps and valves. The
proposed changes do not impact any accident initiators or analyzed
events or assumed mitigation of accident or transient events. They
do not involve the addition or removal of any equipment, or any
design changes to the facility. Therefore, the proposed changes do
not represent 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 do not involve a modification to the
physical configuration of the plant. There is no new equipment to be
installed or a change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new
[[Page 4773]]
accident initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in the types or increases in the
amounts of any effluent that may be released off-site and there is
no increase in individual cumulative occupational exposure.
Therefore, the proposed change does not create the possibility of an
accident of a different kind than previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment revises TS 5.5.6, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves.
The proposed amendment incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves. The safety function of the affected pumps and valves will be
maintained. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Russell Gibbs.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: October 6, 2008.
Description of amendment request: The proposed change would remove
work hour controls and/or references to the NRC Generic Letter 82-12
from the administrative control sections of the technical
specifications. On April 17, 2007, the NRC approved a final rule that
amended 10 CFR Part 26 and, among other changes, established
requirements for managing worker fatigue at operating nuclear power
plants. Subpart I, ``Managing Fatigue,'' specifically addresses
managing worker fatigue by designating individual break requirements,
work hour limits, and annual reporting requirements. Subpart I was
published in the Federal Register on March 31, 2008 (73 FR 16966), with
a required implementation period of 18 months. Compliance is,
therefore, required by October 1, 2009. In order to support compliance
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove
these work hour controls from Technical Specification 5.2.2.e at the
Crystal River Unit 3 Nuclear Generating Plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes remove TS [technical specification]
controls on working hours for personnel who perform safety related
functions. The TS controls are superseded by the worker fatigue
requirements in 10 CFR Part 26. Removal of the TS requirements will
be performed concurrently with the implementation of the 10 CFR Part
26, Subpart I requirements. The proposed changes do not impact the
physical configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. The proposed changes do
not impact the initiators or assumptions of analyzed events, nor do
they impact the mitigation of accidents or transient events.
Therefore, it is concluded that these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26.
Work hours will continue to be controlled in accordance with NRC
requirements. The new rule allows for deviations from controls to
mitigate or prevent a condition adverse to safety or as necessary to
maintain the security of the facility. This ensures that the new
rule will not restrict work hours and thereby create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter plant configuration, require
that new plant equipment be installed, alter assumptions made about
accidents previously evaluated, add any initiators, or effect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes remove TS controls on working hours for
personnel who perform safety related functions. The TS controls are
superseded by the worker fatigue requirements in 10 CFR Part 26. The
proposed changes do not involve any physical changes to plant or the
manner in which plant systems are operated, maintained, modified,
tested, or inspected. The proposed changes do not alter the manner
in which safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
changes will not result in plant operation in a configuration
outside the design basis. The proposed changes will not adversely
affect systems that respond to safely shut down the plant and to
maintain the plant in a safe shutdown condition.
Removal of plant-specific TS administrative requirements will
not reduce a margin of safety because the requirements in 10 CFR
Part 26 are adequate to ensure that worker fatigue is managed.
Therefore, it is concluded that these changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: December 17, 2008.
Description of amendments request: The proposed change would revise
the Crystal River Unit 3 Improved Technical Specifications
Administrative Controls, Section 5.6, to revise the Inservice Testing
Program to incorporate the Technical Specification Task Force (TSTF)
Standard TS Change Traveler, TSTF-479, Revision 0, ``Changes to Reflect
Revision of 10 CFR 50.55a,'' and TSTF-497, Revision 0, ``Limit
Inservice Testing Program SR 3.0.2 Application to Frequencies of 2
Years or Less.''
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:
4. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change revises the CR-3 [Crystal River Unit 3] ITS
[Improved Technical Specifications], Section 5.6.2.9, ``Inservice
Testing Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and
[[Page 4774]]
valves which are classified as ASME [American Society of Mechanical
Engineers] 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 change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. The proposed change does not involve the addition or removal
of any equipment, or any design changes to the facility. Therefore,
this proposed change does not involve an increase in probability or
consequences of an accident previously evaluated.
5. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change revises the CR-3 ITS, Section 5.6.2.9,
``Inservice Testing Program,'' for consistency with the requirements
of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and
valves which are classified as ASME Code Class 1, Class 2, and Class
3. The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves.
The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or involve a change in the methods governing normal plant
operation. The proposed change will not introduce a new accident
initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in types or increases in the
amounts of any effluents that may be released offsite and there is
no increase in individual or cumulative occupational exposure.
Therefore, the proposed change does not create the possibility of an
accident of a different kind than previously evaluated.
6. Does not involve a significant reduction in a margin of
safety[.]
The proposed change revises the CR-3 ITS, Section 5.6.2.9,
``Inservice Testing Program,'' for consistency with the requirements
of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and
valves which are classified as ASME Code Class 1, Class 2, and Class
3. The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change the methods governing normal plant operation.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves. The safety function of the affected pumps and valves will be
maintained. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: November 4, 2008.
Description of amendment request: The proposed amendments would
make changes to the Technical Specifications to increase the 24 month
test load for the Unit 1 Emergency Diesel Generators (EDGs), D1 and D2,
reduce the monthly test load for the Unit 2 EDGs, D5 and D6, and reduce
the 24 month test loads for the Unit 2 EDGs.
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.
This license amendment request proposes to increase a portion of
the Prairie lsland Nuclear Generating Plant Unit 1 emergency diesel
generator's 24-month test loading, reduce the Unit 2 emergency
diesel generators' monthly test loading which demonstrates Technical
Specification operability and revise the 24-month test to require
the Unit 2 emergency diesel generators to operate for at least 2
hours at 100-110% of the continuous rated loading and the remainder
of the 24-hour test at or above 4000 kW. The proposed test loads
will continue to assure that the emergency diesel generators have
the necessary reliability and availability for the design basis
accidents and station blackout events.
The emergency diesel generators are required to be operable in
the event of a design basis accident coincident with a loss of
offsite power to mitigate the consequences of the accident. They are
also the alternate AC source for a station blackout on the other
Prairie lsland Nuclear Generating Plant unit. The emergency diesel
generators are not accident initiators and therefore these changes
do not involve a significant increase in the probability of an
accident previously evaluated.
The accident analyses assume that at least one safeguards bus is
provided with power either from the offsite sources or the emergency
diesel generators. The Technical Specification changes proposed in
this license amendment request will continue to assure that the
emergency diesel generators have the capacity and capability to
assume their maximum auto-connected loads. Thus, the changes
proposed in this license amendment request do not involve a
significant increase in the consequences of an accident previously
evaluated.
The changes proposed in this license amendment 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.
This license amendment request proposes to increase a portion of
the Prairie Island Nuclear Generating Plant Unit 1 emergency diesel
generator's 24-month test loading, reduce the Unit 2 emergency
diesel generators' monthly test loading which demonstrates Technical
Specification operability and revise the 24-month test to require
the Unit 2 emergency diesel generators to operate for at least 2
hours at 100-110% of the continuous rated loading and the remainder
of the 24-hour test at or above 4000 kW. The proposed test loads
will continue to assure that the emergency diesel generators have
the necessary reliability and availability for the design basis
accidents and station blackout events.
The proposed Technical Specification changes do not involve a
change in the plant design, system operation, or the use of the
emergency diesel generators. The proposed changes require the Unit 1
emergency diesel generators to be tested at increased loads and
allow the Unit 2 emergency diesel generator to be tested at reduced
loads which envelope the required safety function loads. These
revised loads continue to demonstrate the capability and capacity of
the emergency diesel generators to perform their required functions.
There are no new failure modes or mechanisms created due to testing
the emergency diesel generators at the proposed test loading.
Testing of the emergency diesel generators at the proposed test
loadings does not involve any modification in the operational limits
or physical design of plant systems. There are no new accident
precursors generated due to the proposed test loadings.
The Technical Specification changes proposed in this license
amendment do not create the possibility of a new or different kind
of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This license amendment request proposes to increase a portion of
the Prairie Island Nuclear Generating Plant Unit 1 emergency diesel
generator's 24-month test loading, reduce the Unit 2 emergency
diesel generators' monthly test loading which demonstrates Technical
Specification operability and revise the 24-month test to require
the Unit 2 emergency diesel generators to operate for at least 2
hours at 100-110% of the continuous rated loading and the remainder
of the 24-hour test at or above 4000 kW. The proposed test loads
will continue to assure that the emergency diesel generators have
the necessary reliability and availability for the design basis
accidents and station blackout events.
[[Page 4775]]
The proposed Technical Specification changes will continue to
demonstrate that the emergency diesel generators meet the Technical
Specification definition of operability, that is, the proposed tests
will demonstrate that the emergency diesel generators will perform
their safety function and the necessary emergency diesel generator
attendant instrumentation, controls, cooling, lubrication and other
auxiliary equipment required for the emergency diesel generators to
perform their safety function loads are also tested at these
proposed loadings. The proposed testing will also continue to
demonstrate the capability and capacity of the emergency diesel
generators to supply their required loss of offsite power loads
coincident with station blackout loads from the opposite unit. Since
the proposed surveillance testing will continue to demonstrate
operability, and the capability and capacity to supply their
required loss of offsite power coincident with opposite unit station
blackout loads, the proposed Technical Specification changes do not
involve a significant reduction in a margin of safety.
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 requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: March 27, 2008, as supplemented by a
letter December 19, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) requirements related to
control building envelope habitability in TS Section 3.7.3 Control Room
Emergency Ventilation (CREV) System, and add TS Section 5.5.13, Control
Building Envelope Habitability Program, to the Administrative Section
of the TSs. The licensee has included conforming technical changes to
the TS Bases. The proposed revision to the Bases also includes
editorial and administrative changes to reflect applicable changes to
the corresponding TS Bases, which were made to improve clarity, conform
to the latest information and references, correct factual errors, and
achieve more consistency with the standard TS NUREGs. The proposed
revision to the TS and associated Bases is similar to the TSTF-448,
Revision 3. The supplement contains additional information related to
smoke and chemical effects and addresses the associated proposed
revision to TS Section 3.7.3, TS Section 5.5.13 and TS Bases 3.7.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed Technical Specification change involve a
significant increase in the probability or consequences of an
accident previously evaluated?
No. The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed Technical Specification change create the
possibility of a new or different kind of accident from any accident
previously evaluated?
No. The proposed change does not impact the accident analysis.
The proposed change does not alter the required mitigation
capability of the CRE emergency ventilation system, or its
functioning during accident conditions as assumed in the licensing
basis analyses of design basis accident radiological consequences to
CRE occupants. No new or different accidents result from performing
the new surveillance or following the new program. The proposed
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or a
significant change in the methods governing normal plant operation.
The proposed change does not alter any safety analysis assumptions
and is consistent with current plant operating practice. Therefore,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed Technical Specification change involve a
significant reduction in a margin of safety?
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. 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: Thomas H. Boyce.
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of amendment request: December 22, 2008 (TS-463-T).
Description of amendment request: The proposed amendment would, on
a one-time basis, extend several Technical Specification (TS)
surveillance frequencies approximately 45 days.
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 requested action is a one-time extension to the performance
interval of a limited number of TS surveillance requirements. The
performance of these surveillances, or the failure to perform these
surveillances, is not a precursor to an accident. Performing these
surveillances or
[[Page 4776]]
failing to perform these surveillances does not affect the
probability of an accident. Therefore, the proposed delay in
performance of the surveillance requirements in this amendment
request does not increase the probability of an accident previously
evaluated.
A delay in performing these surveillances does not result in a
system being unable to perform its required function. In the case of
this one-time extension request, the relatively short period of
additional time that the systems and components will be in service
before the next performance of the surveillance will not affect the
ability of those systems to operate as designed. Therefore, the
systems required to mitigate accidents will remain capable of
performing their required function. No new failure modes have been
introduced because of this action and the consequences remain
consistent with previously evaluated accidents. Therefore, the
proposed delay in performance of the surveillance requirements in
this amendment request does not involve a significant increase in
the consequences of an accident.
Therefore, operation of the facility in accordance with the
proposed license amendment would 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 does not involve a physical alteration of
any system, structure, or component (SSC) or a change in the way any
SSC is operated. The proposed amendment does not involve operation
of any SSCs in a manner or configuration different from those
previously recognized or evaluated. No new failure mechanisms will
be introduced by the one-time surveillance requirement extensions
being requested.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment is a one-time extension of the
performance interval of a limited number of TS surveillance
requirements. Extending these surveillance requirements does not
involve a modification of any TS Limiting Conditions for Operation.
Extending these surveillance requirements does not involve a change
to any limit on accident consequences specified in the license or
regulations. Extending these surveillance requirements does not
involve a change to how accidents are mitigated or a significant
increase in the consequences of an accident. Extending these
surveillance requirements does not involve a change in a methodology
used to evaluate consequences of an accident. Extending these
surveillance requirements does not involve a change in any operating
procedure or process.
The instrumentation and components involved in this request have
exhibited reliable operation based on the results of the most recent
performance of their 24-month surveillance requirements.
Based on the limited additional period of time that the systems
and components will be in service before the surveillances are next
performed, as well as the operating experience that these
surveillances are typically successful when performed, it is
reasonable to conclude that the margins of safety associated with
these surveillance requirements will not be affected by the
requested extension.
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: Thomas Boyce.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: August 1, 2008, as supplemented November
25 and December 31, 2008 (2 letters).
Description of amendment request: The proposed amendment would
revise the following: (1) Technical Specification (TS) 4.2.1, ``Fuel
Assemblies,'' and TS Surveillance Requirements 3.5.1.4,
``Accumulators,'' and 3.5.4.3, ``RWST [Refueling Water Storage Tank],''
to increase the maximum number of Tritium Producing Burnable Absorber
Rods (TPBARs) that can be irradiated per cycle from 400 to 704.
An application that addressed similar issues was previously
submitted on August 1, 2008, and notice of that application was
provided in the Federal Register on November 12, 2008 (73 FR 66946).
Due to certain changes in the specifics of the December 31, 2008,
revision from those proposed in the August 1, 2008, application, as
supplemented on November 25 and December 31, 2008, the application is
being renoticed in its entirety. This notice supersedes the notice
published in the Federal Register on November 12, 2008.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the maximum number of TPBARs in the
core. The required boron concentration for the cold leg accumulators
(CLAs) and RWST remains unchanged. The current boron concentration
has been demonstrated to maintain the required accident mitigation
safety function for the CLAs and RWST with the higher number of
TPBARs and this will be verified for each core that contains TPBARs
as part of the normal reload analysis. The CLAs and RWST safety
function is to mitigate accidents that require the injection of
borated water to cool the core and to control reactivity. These
functions are not potential sources for accident generation and the
modification of the number of TPBARs will not increase the potential
for an accident. Therefore, the possibility of an accident is not
increased by the proposed changes. The current boron concentration
levels are supported by the proposed number of TPBARs in the core.
Since the current boron concentration levels will continue to
maintain the safety function of the CLAs and RWST in the same manner
as currently approved, the consequences of an accident are not
increased by the proposed changes.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change only modifies the maximum number of TPBARs
in the core. The boron concentrations for accident mitigation
functions of the CLAs and RWST remain unchanged. These functions do
not have a potential to generate accidents as they only serve to
perform mitigation functions associated with an accident. The
proposed modification will maintain the mitigation function in an
identical manner as currently approved. There are no plant equipment
or operational changes associated with the proposed revision.
Therefore, since the CLA and RWST functions are not altered and the
plant will continue to operate without change, the possibility of a
new or different kind of an accident is not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This change proposes a change to the maximum number of TPBARs in
the core. The boron concentration requirements that support the
accident mitigation functions of the CLAs and RWST remain unchanged.
The proposed change does not alter any plant equipment or components
and does not alter any setpoints utilized for the actuation of
accident mitigation system or control functions. The proposed number
of TPBARs, in conjunction with the current boron concentration
values, has been demonstrated to provide an adequate level of
reactivity control for accident mitigation and this will be verified
for each core that contains TPBARs as part of the normal reload
analysis. Therefore, the proposed change will 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
[[Page 4777]]
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 Acting Branch Chief: P. Milano.
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.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: November 25, 2008.
Brief description of amendment request: The proposed amendment
would revise Appendix A, Technical Specifications (TS), as they apply
to the spent fuel pool (SFP) storage requirements in TS section 3.7.16
and the criticality requirements for the Region I SFP and north tilt
pit fuel storage racks, in TS section 4.3.1.1.
Date of publication of individual notice in Federal Register:
January 2, 2009 (74 FR 123).
Expiration date of individual notice: February 3, 2009.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: December 1, 2008.
Description of amendment request: By letter dated October 31, 2008,
the Nuclear Regulatory Commission issued Amendment No. 186, to Callaway
Plant, Unit 1, Facility Operating License No. NPF-30. The amendment
allowed a one-time extension of the allowed outage time (completion
time) for each of the two essential service water (ESW) trains (ESW
Train A and Train B) from 72 hours to 14 days. The extended completion
time was requested to support planned replacement of the underground
carbon steel piping with new high density polyethylene (HDPE) piping
for ESW Train A and ESW Train B during plant operation. The amendment
was issued with a requirement to complete the replacement of carbon
steel piping with HDPE piping for both ESW trains by December 31, 2008.
By its application dated December 1, 2008, the licensee informed NRC
that it had experienced significant delays in completing the
replacement of underground piping/conduit due, in part, to underground
obstructions during excavation, a longer refueling outage (Refuel 16)
than anticipated, a forced outage at the beginning of Cycle 17,
switchyard maintenance, and other equipment and personnel issues.
However, the replacement of ESW Train A carbon steel piping was
completed by the required date of December 31, 2008, but the
replacement of ESW Train B carbon steel piping was deferred.
Consequently, the licensee proposed to extend the implementation date
for completion of replacement of carbon steel piping for ESW Train B
from December 31, 2008, to April 30, 2009.
Date of publication of individual notice in Federal Register:
December 23, 2008 (73 FR 78858).
Expiration date of individual notice comment period: January 22,
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
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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendments: June 20, 2008.
Brief description of amendments: The amendments conform the
licenses to reflect the direct transfer of AmerGen Energy Company,
LLC's ownership and operating authority for Clinton Power Station, Unit
No. 1, Oyster Creek Nuclear Generating Station (Oyster Creek), and
Three Mile Island Nuclear Station, Unit 1, to Exelon Generation
Company, LLC, (ECG) as approved by Commission Order dated December 23,
2008. Transfer of the license for Oyster Creek will also authorize EGC
to store spent fuel in the Oyster Creek independent spent fuel storage
installation.
Date of issuance: January 8, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
[[Page 4778]]
Amendment Nos.: CPS-183, Oyster Creek-271, and TMI-1-267.
Facility Operating License Nos. NPF-62, DPR-16, and DPR-50: The
amendments revised the Technical Specifications and Licenses.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50368). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 23, 2008.
No significant hazards consideration comments received: The NRC
received three comments on August 27, 2008, one for each plant's
initial notice. The comments did not provide any information additional
to that in the application, nor did they provide any information
contradictory to that provided in the application.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: April 4, 2008.
Brief description of amendment: The amendment revised the Technical
Specifications by removing the operability and surveillance
requirements for the shield building ventilation (SBV) and auxiliary
building special ventilation filter train heaters, and reducing the
operating time required to verify the SBV system operability from 10
hours to 15 minutes.
Date of issuance: December 30, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 201.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 3, 2008 (73 FR
31720) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 30, 2008.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: April 14, 2008, as supplemented
by letter dated October 17, 2008.
Brief description of amendment: The amendment adds a new footnote
to Kewaunee Technical Specifications Table 3.5-4, ``Instrument
Operating Conditions for Isolation Functions.'' The new footnote allows
the main steam line isolation circuitry to be inoperable when both main
steam isolation valves are closed and deactivated.
Date of issuance: January 12, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 202.
Facility Operating License No. DPR-43: Amendment revised the
operating license and Technical Specifications.
Date of initial notice in Federal Register: June 17, 2008 (73 FR
34340) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 12, 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: December 11, 2007, as
supplemented December 18, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications sections to allow the bypass test times and
Completion Times (CTs) for Limiting Condition for Operation (LCOs)
3.3.1, ``Reactor Trip System (RTS) Instrumentation;'' 3.3.2,
``Engineered Safety Feature Actuation System (ESFAS) Instrumentation;''
3.3.6, ``Containment Air Release and Addition Isolation
Instrumentation,'' and 3.3.9, ``Boron Dilution Mitigation System
(BDMS).''
The proposed license amendment request (LAR) adopts changes as
described in Westinghouse Commercial Atomic Power (WCAP) topical report
WCAP-14333-P-A, Revision 1, ``Probabilistic Risk Analysis of the
Reactor Protection System and Engineered Safety Features Actuation
System Test Times and Completion Times,'' issued October 1998 and
approved by U.S. Nuclear Regulatory Commission (NRC) letter dated July
15, 1998. Implementation of the proposed changes is consistent with
Technical Specification Task Force (TSTF) Traveler TSTF-418, Revision
2, ``RPS [Reactor Protection System] and ESFAS Test Times and
Completion Times (WCAP-14333).'' The NRC approved TSTF-418, Revision 2,
by letter dated April 2, 2003.
In addition, the proposed LAR adopts changes as described in WCAP-
15376-P-A, Revision 1, ``Risk-Informed Assessment of the RTS and ESFAS
Surveillance Test Intervals and Reactor Trip Breaker Test and
Completion Times,'' issued March 2003, as approved by NRC letter dated
December 20, 2002. Implementation of the proposed changes is consistent
with TSTF Traveler TSTF-411, Revision 1, ``Surveillance Test
Interval Extension for Components of the Reactor Protection System
(WCAP-15376).'' The NRC approved TSTF-411, Revision 1, by letter dated
August 30, 2002. The licensee also requested additional changes not
specifically included in the above topical reports. These changes will
be evaluated in a future amendment.
Date of issuance: December 22, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 247 and 240.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 25, 2008 (73 FR
15783). The supplement dated December 18, 2008, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 22, 2008.
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: December 11, 2007, as
supplemented by letter dated December 18, 2008.
Brief description of amendments: The amendments revised the
Technical Specification sections to allow the bypass test times and
Completion Times for Limiting Condition for Operation 3.3.1, ``Reactor
Trip System (RTS) Instrumentation'' and 3.3.2, ``Engineered Safety
Feature Actuation System (ESFAS) Instrumentation.''
By letter dated December 30, 2008 (Agencywide Documents Access and
Management System Accession No. ML0083460216), the NRC issued Amendment
No. 247 and Amendment No. 240 for Catawba Units 1 and 2, respectively,
for all the proposed changes approved by the NRC in TSTFs 411 and 418.
The December 30, 2008, amendment stated that the following changes
would be evaluated in a future amendment:
Surveillance requirement (SR) 3.3.1.5, Safety injection input from
ESFAS, Condition J, Feedwater isolation with low average core
temperature coincident with reactor trip P-4, SR 3.3.2.2, turbine
[[Page 4779]]
trip and feedwater isolation for steam generator water level high high.
(P-14), SR 3.3.2.4 turbine trip and feedwater isolation for steam
generator water level high high (P-14), and SR 3.3.2.5 turbine trip and
feedwater isolation for low average core temperature trip coincident
with reactor trip P-4.
This amendment approves the above changes.
Date of issuance: January 9, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 248 and 241.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 25, 2008 (73 FR
15783). The supplement dated December 18, 2008, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 9, 2009.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: February 6, 2008, as
supplemented by letter dated July 29, 2008.
Brief description of amendment: The amendment revised the
Surveillance Requirements (SRs) for control rod exercising from weekly
to monthly in Technical Specification (TS) 4.3.A.2, revise verification
of control rod coupling integrity as described in TS 4.3.B.1, revise
the scram insertion time Limiting Conditions for Operation (LCOs) and
SRs as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D,
the LCO and SR for Control Rod Accumulators.
Date of issuance: January 7, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 233.
Facility Operating License No. DPR-28: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: March 11, 2008 (73 FR
13024). The supplemental letter dated July 29, 2008, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated January 7, 2009.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: July 30, 2008, as supplemented by letter
dated October 2, 2008.
Brief description of amendment: The amendment revises the current
TS 3.6.6.3 surveillance requirements for sodium hydroxide (NaOH)
concentration. Specifically, the amendment changes the surveillance
requirements of the NaOH tank solution concentration from between 5.0
weight (wt.) percent and 16.5 wt. percent to between 6.0 wt. percent
and 8.5 wt. percent.
Date of issuance: January 13, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1--234.
Renewed Facility Operating License No. DPR-51: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: November 4, 2008, (73
FR 65694). The supplement dated October 2, 2008, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 13, 2009.
No significant hazards consideration comments received: No.
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, Claiborne
County, Mississippi
Date of application for amendment: June 30, 2008.
Brief description of amendment: The amendment (1) deleted Technical
Specification (TS) surveillance requirement (SR) 3.1.3.2 and revised SR
3.1.3.3; (2) removed the reference to SR 3.1.3.2 from Required Action
A.2 of TS 3.1.3, ``Control Rod OPERABILITY''; (3) clarified the
requirement to fully insert all insertable rods for the limiting
condition for operation in TS 3.3.1.2 Required Action E.2, ``Source
Range Monitoring Instrumentation''; and (4) revised Example 1.4-3 in
Section 1.4, ``Frequency,'' to clarify the applicability of the 1.25
surveillance test interval extension.
Date of issuance: December 31, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 180.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50359).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 31, 2008.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3, Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: January 17, 2008.
Brief description of amendment: The amendment revises the Crystal
River, Unit 3 Improved Technical Specification Surveillance Requirement
3.7.5.2, ``Emergency Feedwater System,'' to align the text for the
emergency feedwater system surveillance frequency with the text in the
Technical Specifications Task Force Standard Technical Specification
Change Traveler-101, Revision 0 and the NRC technical report, NUREG-
1430, Volume 1, Revision 3, ``Standard Technical Specifications Babcock
and Wilcox Plants--Specification.''
Date of issuance: January 9, 2009.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 231.
Facility Operating License No. DPR-72: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: May 20, 2008 (73 FR
29163).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 9, 2009.
No significant hazards consideration comments received: No.
[[Page 4780]]
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 application for amendments: December 17, 2007, as
supplemented by letters dated October 2, and November 18, 2008.
Brief description of amendments: The amendments increase the
completion times (CTs) for required actions related to Technical
Specifications (TS) 3.5.2, regarding the Emergency Core Cooling System,
and 3.6.6, regarding the Containment Spray and Cooling Systems from 72
hours to 14 days. In addition, invalid notes were deleted from TSs
3.5.2 and 3.6.6 and new notes were added to specify the limitations on
the use of the 14-day extended CT.
Date of issuance: December 31, 2008.
Effective date: As of its date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: Unit 1--202; Unit 2--203.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: January 29, 2008 (73 FR
5227). The supplement(s) dated October 2 and November 18, 2008,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 31, 2008.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: July 7, 2008.
Brief description of amendments: The amendments revised the
Technical Specification (TS) testing frequency for the Surveillance
Requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The change
revised the frequency of SR 3.1.4.2, control rod scram time testing,
from ``120 days cumulative operation in Mode 1'' to ``200 days
cumulative operation in Mode 1.'' These changes are based on TS Task
Force (TSTF) change traveler TSTF-460 (Revision 0) that has been
approved generically for the Boiling-Water Reactor (BWR) Standard TS,
NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency of
SR 3.1.4.2, control rod scram time testing, from ``120 days cumulative
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
Date of issuance: January 2, 2009.
Effective date: January 2, 2009.
Amendment Nos.: 249 for Unit 1 and 228 for Unit 2
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: October 7, 2008 (73 FR
58675).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 2, 2009.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: July 7, 2008.
Brief description of amendments: The amendment adopted the Nuclear
Regulatory Commission (NRC) approved Technical Specification Task Force
(TSTF) change traveler TSTF-475, (Revision 1), ``Control Rod Notch
Testing Frequency and SRM [Source Range Monitor] Insert Control Rod
Action,'' to change the Standard Technical Specifications (STS) for
General Electric (GE) Plants (NUREG-1433, BWR/4 to the plant-specific
TS, that allows: (1) Revising the frequency of Surveillance Requirement
(SR) 3.1.3.2, notch testing of fully withdrawn control rod, from ``7
days after the control rod is withdrawn and THERMAL POWER is greater
than the LPSP of RWM'' to ``31 days after the control rod is withdrawn
and THERMAL POWER is greater than the LPSP [Low Power Set Point] of the
RWM [Rod With Minimizer]'', and (2) revising Example 1.4-3 in Section
1.4 ``Frequency'' to clarify that the 1.25 surveillance test interval
extension in SR 3.0.2 is applicable to time periods discussed in NOTES
in the ``SURVEILLANCE'' column in addition to the time periods in the
``FREQUENCY'' column.
Date of issuance: January 2, 2009.
Effective date: January 2, 2009.
Amendment Nos.: 250 for Unit 1 and 229 for Unit 2.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: October 7, 2008 (73 FR
58675).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 2, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: September 19, 2008.
Brief description of amendment: The amendment modifies the Final
Safety Analysis Report by requiring an inspection of the ice condenser
within 24 hours of experiencing a seismic event greater than or equal
to an operating basis earthquake within the 5-week period after ice
basket replenishment has been completed to confirm that adverse ice
fallout has not occurred that could impede the ability of the ice
condenser lower inlet doors to open.
Date of issuance: January 6, 2009.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 73.
Facility Operating License No. NPF-90: Amendment authorizes
revision to the FSAR.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65698).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 6, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: March 27, 2008, as supplemented
September 26, 2008.
Brief description of amendment: The amendment revises the allowable
value listed for Function 3, ``Containment Purge Exhaust Radiation
Monitors,'' in Table 3.3.6-1, ``Containment Vent Isolation
Instrumentation,'' of the limited condition for operation 3.3.6.
Date of issuance: January 8, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 74.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications and License.
Date of initial notice in Federal Register: May 6, 2008 (73 FR
25047). The supplement dated September 26,
[[Page 4781]]
2008, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 8, 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 amendment: December 17, 2007, as
supplemented on July 22, 2008, September 26, 2008, and November 25,
2008.
Brief description of amendment: These amendments revised Technical
Specification (TS) 3.8.3 to allow a one-time extended 14-day completion
time (CT) for each of the two underground diesel fuel oil storage tanks
(FOST) to permit removal of the current coating and to recoat the tanks
in preparation for use of ultra-low sulfur diesel fuel oil. The change
revised the TS to extend the CT associated with an inoperable emergency
diesel generator FOST from 7 days to 14 days, applicable once for each
of the two tanks.
Date of issuance: December 31, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 254 and 235.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: January 15, 2008 (73 FR
2552). The supplements dated July 22, 2008, September 26, 2008, and
November 25, 2008, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated December 31, 2008.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 15th day of January 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-1568 Filed 1-26-09; 8:45 am]
BILLING CODE 7590-01-P