[Federal Register Volume 78, Number 237 (Tuesday, December 10, 2013)]
[Notices]
[Pages 74176-74188]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-29168]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0266]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to 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 or combined
license, as applicable, 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 November 14, 2013 to November 27, 2013. The
last biweekly notice was published on November 26, 2013 (78 FR 70589).
ADDRESSES: You may submit comment by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0266. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: 3WFN, 06-44M, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0266 when contacting the NRC
about the availability of information regarding this document. You may
access publicly-available information related to this action by the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0266.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly-available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. The ADAMS accession number
for each document referenced in this notice (if that document is
available in ADAMS) is provided the first time that a document is
referenced.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2013-0266 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission.
[[Page 74177]]
The NRC posts all comment submissions at http://www.regulations.gov as
well as entering the comment submissions into ADAMS. The NRC does not
routinely edit comment submissions to remove identifying or contact
information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined 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 Section 50.92 of Title 10 of the Code
of Federal Regulations (10 CFR), 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.
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 or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC regulations are accessible electronically from the NRC Library on
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then 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
[[Page 74178]]
accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007).
The E-Filing process requires participants to submit and serve all
adjudicatory documents over the internet, or in some cases to mail
copies on electronic storage media. Participants may not submit paper
copies of their filings unless they seek an exemption in accordance
with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC's guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email to
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC's Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter
[[Page 74179]]
problems in accessing the documents located in ADAMS should contact the
NRC PDR's Reference staff at 1-800-397-4209, 301-415-4737, or by email
to [email protected].
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: September 12, 2013.
Description of amendment request: The proposed amendments revise
technical specification 3.3.2, Emergency Safety Feature Actuation
System (ESFAS) Instrumentation, to support planned plant modifications
associated with NRC Order EA-12-049, Order Modifying Licenses with
Regard to Requirements for Mitigation Strategies for Beyond-Design-
Basis External Events. Specifically, the amendment modifies the
Allowable Value and Nominal Trip Setpoints listed in Table 3.3.2-1,
Function 6.f, Auxiliary Feedwater pump suction transfer on low suction
pressure.
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:
Criterion 1: Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes are in support of a plant modification
involving the installation of an AC-independent AFW Suction Transfer
scheme and hardware to ensure a continuous AFW suction source during
an Extended Loss of AC Power (ELAP) event. The purpose of Table
3.3.2-1 Function 6.f is to preserve the AFW pumps by ensuring a
continuous suction supply to the pumps. The proposed change will
cause the AFW pumps to align to the safety-related suction source
sooner than under the current setpoint values for design basis
events. The result of the proposed TS setpoint changes will be an
increase in margin for AFW pump suction. The new TS setpoints were
selected with sufficient margin for instrument uncertainty to ensure
that the safety-related AFW suction transfer function actuates
before the new AC independent AFW suction transfer function and to
prevent any adverse interaction of the two schemes. In other words,
the proposed change will ensure the safety-related suction transfer
is initiated before the non-safety AC independent AFW suction
transfer initiates. The specific TS changes are associated with 1)
the specific Nominal Trip Setpoint and Allowable Values for the AFW
Pump Suction Transfer on Suction Pressure--Low feature, 2) the
addition of specific requirements to be taken if the as-found
channel setpoint is outside its predefined as-found tolerance, and
3) the addition of specific requirements regarding resetting of an
channel setpoint within an as-left tolerance.
The AFW Pump Suction Transfer on Suction Pressure--Low feature
does not affect the probability of any accident being initiated. In
addition, none of the abovementioned proposed TS changes affect the
probability of any accident being initiated.
Actuation of the AFW Pump Suction Transfer on Suction Pressure--
Low feature will continue to ensure that adequate AFW pump suction
is maintained during design bases events. Transfer to the safety-
related suction source will actually occur earlier due to the
proposed change. The proposed changes to Nominal Trip Setpoints and
Allowable Values are based on accepted industry standards and will
preserve assumptions in the applicable accident analyses. None of
the proposed changes alter any assumption previously made in the
radiological consequences evaluations, nor do they affect mitigation
of the radiological consequences of an accident previously
evaluated.
In summary, the proposed changes will not involve any increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2: Does the proposed amendment reate the possibility of a new
or different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any of the proposed changes.
The AFW Pump Suction Transfer feature is not an accident initiator.
No changes to the overall manner in which the plant is operated are
being proposed. Therefore, none of the proposed changes will create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant system
pressure boundary, and the containment barriers. The proposed TS
setpoints serve to ensure proper AFW system suction transfer for
design bases events, whereby the proposed TS changes will not have
any effect on the margin of safety of fission product barriers. In
addition, the proposed TS changes will not have any impact on these
barriers. No accident mitigating equipment will be adversely
impacted as a result of the modification. Therefore, existing safety
margins will be preserved. None of the proposed changes will involve
a significant reduction in a margin of safety.
Based on the above, it is concluded that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Robert J. Pascarelli.
Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit 2, Darlington County, South Carolina
Date of amendment request: September 10, 2013.
Description of amendment request: The proposed change would revise
Technical Specification Limiting Condition for Operation 3.8.1,
Required Action (RA) B.3.2.2, ``One DG [Diesel Generator] Inoperable--
Perform SR [Surveillance Requirement] 3.8.1.2 for OPERABLE DG within 96
hours,'' by a NOTE clarifying RA B.3.2.2 that states, ``Not required to
be performed when the cause of the inoperable DG is pre-planned
maintenance and testing.''
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 eliminates a conditional surveillance of the
Operable EDG [emergency diesel generator] whenever the alternate
division EDG is out of service for pre-planned maintenance and
testing. The EDG are [is] not an initiator of any accident
previously evaluated. As a result, the probability of any accident
previously evaluated is not significantly increased.
The consequences of any accident previously evaluated are not
increased, as the EDG will continue to meet its safety function to
supply backup AC [alternating current] power as specified in the
accident analysis, in a highly reliable manner, as a common cause
problem between the two EDGs will have been precluded, the alternate
division EDG will no longer be taken out of service for testing, and
its normally scheduled surveillances will be met.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 74180]]
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. The changes do not alter assumptions made in the safety
analysis for EDG performance.
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 eliminates a conditional surveillance of the
Operable EDG whenever the alternate division EDG is out of service
for pre-planned maintenance and testing. The EDG will continue to
meet its specified safety function in the safety analysis to provide
backup AC power, in a highly reliable manner, as a common cause
problem between the two EDGs will have been precluded, the alternate
division EDG will no longer be taken out of service for testing, and
its normally scheduled surveillances will be met.
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: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon Street, Charlotte, NC 28202.
NRC Branch Chief: Jessie F. Quichocho.
Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit 2, Darlington County, South Carolina
Date of amendment request: September 30, 2013.
Description of amendment request: The proposed amendment implements
the Nuclear Regulatory Commission (NRC)-approved Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-491, ``Removal of Main Steam and Main Feedwater Valve
Isolation Times from Technical Specifications,'' via the Consolidated
Line Item Improvement Process (CLIIP). This request will modify the
current Unit 2 Technical Specifications (TSs) 3.7.2, Main Steam
Isolation Valves and 3.7.3, Main Feedwater Isolation Valves, Main
Feedwater Regulation Valves and Bypass Valves by relocating the
specific isolation time for the isolation valves from the associated
Surveillance Requirements (SRs). The isolation time in the TS SRs is
replaced with the requirement to verify the valve isolation time is
``within limits.'' The specific isolation times will be maintained in
the Unit 2 Technical Requirements Manual.
The NRC staff published a notice of opportunity for comment in the
Federal Register on October 5, 2006 (71 FR 58884), on possible
amendments adopting TSTF-491, Revision 2, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the CLIIP. The NRC staff subsequently issued a
notice of availability of the models for referencing in license
amendment applications in the Federal Register on December 29, 2006 (71
FR 78472). The licensee affirmed the applicability of the following
NSHC determination in its application dated September 30, 2013.
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 is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows relocating main steam and main
feedwater valve isolation times to the Licensee Controlled Document
that is referenced in the Bases. The proposed change is described in
Technical Specification Task Force (TSTF) Standard TS Change
Traveler TSTF-491 related to relocating the main steam and main
feedwater valves isolation times to the Licensee Controlled Document
that is referenced in the Bases and replacing the isolation time
with the phase, ``within limits.''
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed changes relocate the main steam and main feedwater
isolation valve times to the Licensee Controlled Document that is
referenced in the Bases. The requirements to perform the testing of
these isolation valves are retained in the TS. Future changes to the
Bases or licensee-controlled document will be evaluated pursuant to
the requirements of 10 CFR 50.59, ``Changes, test and experiments,''
to ensure that such changes do not result in more than minimal
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological consequences of any accident previously
evaluated. Further, the proposed changes do not increase the types
and the amounts of radioactive effluent that may be released, nor
significantly increase individual or cumulative occupation/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
Criterion 2: The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed changes relocate the main steam and main feedwater
valve isolation times to the Licensee Controlled Document that is
referenced in the Bases. In addition, the valve isolation times are
replaced in the TS with the phase ``within limits.'' The changes do
not involve a physical altering of the plant (i.e., no new or
different type of equipment will be installed) or a change in
methods governing normal pant operation. The requirements in the TS
continue to require testing of the main steam and main feedwater
isolation valves to ensure the proper functioning of these isolation
valves.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed changes relocate the main steam and main feedwater
valve isolation times to the Licensee Controlled Document that is
referenced in the Bases. In addition, the valve isolation times are
replaced in the TS with the phase ``within limits.'' Instituting the
proposed changes will continue to ensure the testing of main steam
and main feedwater isolation valves. Changes to the Bases or license
controlled document are performed in accordance with 10 CFR 50.59.
This approach provides an effective level of regulatory control and
ensures that main steam and feedwater isolation valve testing is
conducted such that there is no significant reduction in the margin
of safety.
The margin of safety provided by the isolation valves is
unaffected by the proposed changes since there continue to be TS
requirements to ensure the testing of main steam and main feedwater
isolation valves. The proposed changes maintain sufficient controls
to preserve the current margins of safety.
The NRC staff proposes to determine that the amendment request
involves NSHC.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon Street, Charlotte, NC 28202.
[[Page 74181]]
NRC Branch Chief: Jessie F. Quichocho.
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: July 29, 2013.
Description of amendment request: The amendment would add a
permanent exception to the River Bend Station (RBS) Technical
Requirements Manual (TRM) Section 3.9.14, ``Crane Travel--Spent and New
Fuel Storage, Transfer, and Upper Containment Fuel Pools,'' to allow
for movement of fuel pool gates over fuel assemblies for maintenance.
This exception will also be described by revision to the RBS Updated
Safety Analysis Report (USAR) Section 9.1.2.2.2, ``Fuel Building Fuel
Storage,'' and Section 9.1.2.3.3, ``Protection Features of Spent Fuel
Storage Facilities.''
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. Involved a significant increase in the probability or
consequences of an accident previously evaluated.
Response: No.
The RBS fuel building fuel storage facilities consist of three
interconnected stainless steel-lined concrete pools. The spent fuel
storage pool is the largest of these pools. Adjacent to the fuel
storage pool are the cask pool and the lower IFTS [inclined fuel
transfer system] pool. Each of these two pools is separated from the
fuel storage pool by a full-height wall encompassing a watertight
gate. The watertight gates are normally open, but are closed to seal
their respective pools during cask handling and equipment
maintenance operations. It is necessary to lift the gates from the
pools for maintenance or seal replacement. The total weight of the
gate including the rigging equipment is 2000 pounds. This lift is
considered as a heavy load lift since it is higher than the current
analyzed light load limit of 1200 pounds for movement of loads over
fuel assemblies. TRM 3.9.14 prohibits any load in excess of 1200
pounds from travel over fuel assemblies in the storage pool.
Each of the gates is designed with a pneumatic seal that, when
pressurized, seals the respective pool from the spent fuel pool,
forming a watertight barrier. No provisions for moving the gates
over fuel assemblies were included in the current licensing basis
for RBS heavy loads. However, the service life qualification of the
gate seals necessitates that they be replaced several times over the
life of the plant. Therefore, approval of an exception to the
current prohibition is required to allow for replacement of the gate
seals.
To perform the movement of the gate from its installed position
to a position where the seal can be replaced, an engineering plan
that meets the intent of the applicable regulatory guidance has been
developed. RBS' program for control of heavy load movements complies
with that guidance, and this will prevent the gate from dropping
onto the spent fuel assemblies during the movement activity. The
program features include the design of the lifting devices, design
of the cask and fuel bridge cranes, crane operator training, and the
use of written procedures. The regulatory guidance will be met in
all respects, except that, in lieu of a single failure-proof crane,
the method will employ redundant and diverse means to meet the
intent of single-failure proof movements.
Entergy proposes to lift the spent fuel pool gate using a
rigging method that complies with the intent of the guidance of
References 10.c through 10.f [of the licensee's letter dated July
29, 2013]. The proposed method will be accomplished through the use
of fuel building bridge crane and the cask crane at the same time to
provide the redundancy required to make the lift single-failure
proof and satisfy single-failure proof criteria.
In the proposed method, the fuel building bridge crane and the
cask crane will be used to perform the gate lifting and movement.
The intent of the applicable regulatory guidance is that in lieu of
providing a single-failure-proof crane system, the control of heavy
loads guidelines can be satisfied by establishing that the potential
for a heavy load drop is extremely small. The gate lifting using the
bridge crane and cask crane will conform to applicable regulatory
guidelines, in that the probability of the gate drop over the spent
fuel assemblies is extremely small. Both cranes have a rated
capacity of 15 tons. The maximum weight of the gate and rigging is
2000 pounds. Therefore, there is ample safety factor margin for
lifting and movements of the subject spent fuel pool gate. Special
lifting devices, which have redundancy or ultimate strength of at
least ten times the lifted load, will also be utilized during the
rigging process. Even though neither the fuel building bridge crane
or the cask crane is a single-failure proof crane, rigging the spent
fuel pool gate using both cranes will provide the required
redundancy that meets the intent of single-failure proof criteria.
The proposed load lift of the fuel pool gate for replacement of
the seal conforms to all of the applicable regulatory guidelines.
The design of the lifting lugs and associated rigging (e.g., chains,
slings, shackles, hoists, etc.) conforms to the guidelines of NUREG-
0612, [``Control of Heavy Loads at Nuclear Power Plants,''] Section
5.1.6, and ``Single-Failure Proof Handling System,'' and References
10.d through 10.f [of the licensee's letter dated July 29, 2013].
The auxiliary hook of the cask crane has a rated capacity of 15
tons. The cask crane is not a single-failure-proof crane. However,
it meets NUREG-0612 criteria of Section 5.1.1(6) and is designed for
seismic loading. As discussed above, the cask crane, alone, will
handle the gate only after the gate is located inside the cask pool
where drop of the gate above the spent fuel rack is no longer a
concern. The cask pool area has been evaluated for an accidental
drop of the spent fuel cask. There is no safety-related equipment
inside the cask pool. The analyzed maximum weight of the gate and
rigging is 2500 pounds. Therefore, there is ample safety factor
margin for lifting the gate with the cask crane.
The probability and consequences of a seismic event are not
affected by the proposed gate lift. The consequences of a seismic
event during the gate lifting are insignificant since both cranes,
the fuel building bridge crane and the cask crane, are seismically
qualified for the lifted load. In addition, the design of all
rigging conforms to NUREG-0612 guidelines, with a safety factor of
10 for the weight of the load.
Consistent with the defense-in-depth approach outlined in the
guidance, the movement will be conducted according to load handling
instructions. Operator training will be conducted on the activity
prior to the movement, and the equipment will be inspected before
the movement will be performed. NUREG-0612 gives guidance that when
a particular heavy load must be brought over spent fuel, alternative
measures may be used. The combination of preventative measures, as
proposed, minimizes the risks inherent in hauling large loads over
spent fuel to permissible levels. Considering these provisions and
the applicable regulatory guidance, the increase in probability of a
load drop is negligible.
It is therefore concluded that the proposed gate lifting and
movement does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Response: No.
The lifting of the fuel pool gate in the spent fuel pool as
described above minimizes the possibility of a heavy load drop onto
spent fuel assemblies as not credible in accordance with single-
failure-proof criteria. In addition, movement of the gate in the
cask pool using the cask crane does not create the possibility of a
new or different kind of accident. The cask drop accident scenario
in the current RBS licensing basis (since the cask crane is not a
single-failure-proof crane) envelops the accidental drop of the gate
in the cask pool during handling by the cask crane. The analyzed
weight of a cask is 125 tons, as compared to the 1 ton combined
weight of the gate and the rigging.
It is therefore concluded that the proposed gate lifting does
not create the possibility of a new or different kind of accident
from any previously analyzed.
3. Invoke a significant reduction in a margin of safety.
Response: No.
By following the guidance of References 10.c through 10.f [of
the licensee's letter dated July 29, 2013], the movement of the
spent fuel pool gates will have no impact on the analyses of
postulated design basis events for RBS. The NRC guidance provides an
acceptable means of ensuring the appropriate level of safety and
protection against load drop accidents. Therefore, there is no
reduction in the margin of safety associated
[[Page 74182]]
with postulated design basis events at RBS in allowing the proposed
change to the RBS licensing basis. RBS will continue to meet its
commitment to comply with the applicable guidance.
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: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: September 5, 2013.
Description of amendment request: The proposed amendments would
revise Technical Specification 5.5.13, ``Primary Containment Leakage
Rate Testing Program,'' to increase the peak calculated primary
containment internal pressure, Pa, from 39.9 psig to 42.6
psig. The proposed increase in Pa reflects a lower initial
drywell temperature and a number of other modeling changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided on
September 5, 2013, its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to Pa does not alter the assumed
initiators to any analyzed event. The probability of an accident
previously evaluated will not be increased by this proposed change
since this change does not modify the plant or how it is operated.
The change in Pa will not affect radiological dose
consequence analyses. LSCS radiological dose consequence analyses
are based on the maximum allowable containment leakage rate. Even
though the test pressure at which leak rate testing is performed is
Pa, the maximum allowable containment leakage rate is
defined in terms of a percentage of weight of the original content
of containment air, which is independent of the peak calculated
primary containment internal pressure. The Appendix J containment
leak rate testing program will continue to ensure that containment
leakage remains within the leakage assumed in the offsite dose
consequence analyses. The consequences of an accident previously
evaluated will not be increased by this proposed change.
Therefore, operation of the facility in accordance with the
proposed change to Pa will 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 provides a higher Pa than
currently described in the TS. This change is the result of a LOCA-
Drywell Temperature sensitivity analysis performed by General
Electric Hitachi. The peak calculated primary containment internal
pressure remains below the containment design pressure of 45 psig.
This change does not involve any alteration in the plant
configuration (no new or different type of equipment will be
installed) or make changes in the methods governing normal plant
operation. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed change to TS 5.5.13 would 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 peak calculated primary containment internal pressure
remains below the containment design pressure of 45 psig. LSCS
radiological consequence analyses are based on the maximum allowable
containment leakage rate. The change in the peak calculated primary
containment internal pressure does not represent a significant
change in the margin of safety. Operation of the facility in
accordance with the proposed change to TS 5.5.13 does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Tamra Domeyer, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: September 20, 2013.
Description of amendment request: The proposed amendments would
revise Technical Specification 3.3.8.1-1, ``Loss of Power
Instrumentation,'' Table 1, to change the allowable values to address
non-conservative assumptions. The proposed change involves revising the
surveillance requirements to modify the allowable values for the 4.16
kV emergency buses during loss of voltage testing and calibration to
ensure that existing design requirements remain satisfied.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided on
September 20, 2013, its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the 4.16 kV [engineered safety functions]
ESF bus loss of voltage allowable values allow the protection scheme
to function as originally designed. (This change will involve
alteration of nominal trip setpoints in the field and will also be
reflected in revisions to the calibration procedures.) The proposed
change does not affect the probability or consequences of any
accident. Analysis was conducted and demonstrates that the proposed
allowable values will allow the normally operating safety-related
motors to continue to operate without sustaining damage or tripping
during the worst-case, non-accident degraded voltage condition for
the maximum possible time-delay of 5.7 minutes. Thus, these safety-
related loads will be available to perform their safety function if
a loss-of-coolant accident (LOCA) concurrent with a loss-of-offsite
power (LOOP) occurs following the degraded voltage condition.
The proposed changes do not adversely affect accident initiators
or precursors, and do not alter the design assumptions, conditions,
or configuration or the plant or the manner in which the plant is
operated or maintained. The proposed allowable values ensure that
the 4.16 kV distribution system remains connected to the offsite
power system when adequate offsite voltage is available and motor
starting transients are considered. The diesel start due to a LOCA
signal is not adversely affected by this change. During an actual
loss of voltage condition, the loss of voltage time delay will
continue to isolate the 4.16 kV distribution system from offsite
power before the diesel is ready to assume the emergency loads,
which is the limiting time basis for mitigating system responses to
the accident. For this reason, the existing loss of power/LOCA
analysis continues to be valid.
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.
[[Page 74183]]
The proposed change involves the revision of 4.16 kV ESF bus
loss of voltage allowable values to satisfy existing design
requirements. The proposed change does not introduce any changes or
mechanisms that create the possibility of a new or different kind of
accident. The proposed change does not install any new or different
type of equipment, and installed equipment is not being operated in
a new or different manner. No new effects on existing equipment are
created nor are any new malfunctions introduced.
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 protection voltage allowable values are low enough
to prevent inadvertent power supply transfer, but high enough to
ensure that sufficient power is available to the required equipment.
The diesel start due to a LOCA signal is not adversely affected by
this change. During an actual loss of voltage condition, the loss of
voltage time delays will continue to isolate the 4.16 kV
distribution system from offsite power before the diesel is ready to
assume the emergency loads.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Tamra Domeyer, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company (EGC), LLC, Docket Nos. STN 50-456 and STN
50-457, Braidwood Station, Units 1 and 2, Will County, Illinois
Date of amendment request: October 10, 2013.
Description of amendment request: The proposed amendment would
revise the date for the performance of the containment leakage rate
Type A test from ``no later than May 4, 2014,'' to ``prior to entering
MODE 4 at the start of Cycle 18.'' Additionally, EGC is proposing to
establish a requirement for Braidwood Station, Unit 2, to exit the
MODEs of applicability for Containment as described in Technical
Specification 3.6.1, ``Containment'' (i.e., MODEs 1-4), no later than
May 4, 2014.
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:
EGC has evaluated the proposed change for Braidwood Station,
Units 1 and 2 using the criteria in 10 CFR 50.92, and has determined
that the proposed change does not involve a significant hazards
consideration. The following information is provided to support a
finding of no significant hazards consideration.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the Braidwood Station, Units 1 and 2
Containment Leakage Rate Testing Program does not involve a physical
change to the plant or a change in the manner in which the plant is
operated or controlled. The containment function is to provide an
essentially leak tight barrier against the uncontrolled release of
radioactivity to the environment for postulated accidents. As such,
the containment itself, and the testing requirements to periodically
demonstrate the integrity of the containment, exist to ensure the
plant's ability to mitigate the consequences of an accident do not
involve any accident precursors or initiators. Therefore, the
probability of occurrence of an accident previously evaluated is not
significantly increased by the proposed amendment. Implementation of
the proposed change will continue to provide adequate assurance that
during design basis accidents, the containment and its components
would limit leakage rates to less than the values assumed in the
plant safety analyses. Therefore, the consequences of an accident
previously evaluated will not be increased by this proposed change.
Therefore, operation of the facility in accordance with the
proposed administrative change to the date for the performance of
the Unit 2, Type A containment leak rate test will 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 containment, and the testing requirements to periodically
demonstrate the integrity of the containment, exist to ensure the
plant's ability to mitigate the consequences of an accident, and do
not involve any accident precursors or initiators. The proposed
change does not involve a physical change to the plant (i.e., no new
or different type of equipment will be installed) or a change to the
manner in which the plant is currently operated or controlled.
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.
This proposed change does not alter the manner in which safety
limits, limiting safety system setpoints, or limiting conditions for
operation are determined. The specific requirements and conditions
of the containment leakage rate testing program, as proposed, will
continue to ensure that the degree of containment structural
integrity and leak-tightness that is considered in the plant's
safety analysis is maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above evaluation, EGC concludes that the proposed
amendment does not involve a significant hazards consideration under
the standards set forth in 10 CFR 50.92, paragraph (c), and
accordingly, a finding of no significant hazards consideration is
justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: June 6, 2013.
Description of amendment request: The proposed amendment would
change the current requirement that ``each ADS [Automatic
Depressurization System] valve opens when manually actuated,'' to the
requirement that ``each ADS valve actuator strokes when manually
actuated.'' Additionally, the surveillance frequency would change from
``24 months on a STAGGERED TEST BASIS for each valve solenoid,'' to
``24 months.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not modify the method of demonstrating
the operability of the Safety/Relief Valves (S/RVs) in both the
safety and relief modes of operation. The proposed change does
modify the method for demonstrating the proper mechanical
functioning of the S/RVs. The S/RVs are required to function in the
safety mode to prevent overpressurization of the reactor vessel and
reactor coolant system pressure
[[Page 74184]]
boundary during various analyzed transients, including Main Steam
Isolation Valve closure. S/RVs associated with the Automatic
Depressurization System are also required to function in the relief
mode to reduce reactor pressure to permit injection by low pressure
Emergency Core Cooling System (ECCS) pumps during certain reactor
coolant pipe break accidents. The current testing method
demonstrates the proper mechanical functioning of the S/RVs in both
modes through manual actuation of the S/RVs. The proposed testing
method results in acceptable demonstration of the S/RV functions in
both the safety and relief modes, and therefore provides assurance
that the probability of S/RV failure will not increase. None of the
accident safety analyses are affected by the requested [Technical
Specification] TS changes and the consequences of accidents
mitigated by the S/RVs will not increase.
Therefore, the proposed amendment does not result in a
significant increase in the probability or consequences of any
previously evaluated accident.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change modifies the method of testing of the S/RVs,
but does not alter the functions or functional capabilities of the
S/RVs. Testing under the proposed method is performed in offsite
test facilities and in the plant during outage periods when the S/RV
functions are not required. Existing analyses address events
involving an S/RV inadvertently opening or failing to reclose.
Analyses also address the failure of one or more S/RVs to open. The
proposed change does not introduce any new failure mode.
Therefore, it 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 amendment provides for a complete verification of
the functional capability of the S/RVs by performing tests,
inspections, and maintenance activities without opening the valves
while installed in the plant. This alternative testing and
associated programmatic controls will provide an overall level of
assurance that the S/RVs are capable of performing their intended
accident mitigation safety functions. The proposed amendment does
not affect the valve setpoints or adversely affect any other
operational criteria assumed for accident mitigation. No changes are
proposed that alter the setpoints at which protective actions are
initiated, and there is no change to the operability requirements
for equipment assumed to operate for accident mitigation. Moreover,
it is expected that the alternative testing methodology will
increase the margin of safety by reducing the potential for S/RV
leakage resulting from testing. Additionally, the increased testing
frequency of the manual actuation circuitry is beneficial since the
valves will no longer be tested on a staggered test frequency.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
Acting NRC Branch Chief: John G. Lamb.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: June 6, 2013.
Description of amendment request: This proposed change adds a
footnote to Function 6c in Technical Specification Table 3.3.6.1-1.
This change allows only one Trip System to be operable in MODES 4 and 5
for the Manual Initiation Function for Shutdown Cooling System
isolation.
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 manual isolation function of the RHR [Residual Heat Removal]
Shutdown Cooling System is not credited in any FSAR [Final Safety
Analysis Report] safety analysis. The addition of Footnote (c) to
the manual isolation function in TS [Technical Specification] Table
3.3.6.1-1 allows one of the two trip systems to be inoperable in
MODES 4 and 5 and does not alter any equipment.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The addition of Footnote (c) to the manual isolation function in
TS Table 3.3.6.1-1 allows one of the two trip systems to be
inoperable in MODES 4 and 5 and is consistent with other isolation
function required for isolation in MODES 4 and 5.
No new equipment is being introduced, and installed equipment is
not being operated in a new or different manner. There are no set
points, at which protective or mitigative actions are initiated,
affected by this change. These changes do not alter the manner in
which equipment operation is initiated, nor will the function
demands on credited equipment be changed. No alterations in the
procedures that ensure the plant remains within analyzed limits are
being proposed, and no major changes are being made to the
procedures relied upon to respond to an off-normal event as
described in the FSAR. As such, no new failure modes are being
introduced. The proposed change does not alter assumptions made in
the safety analysis and licensing basis since the manual isolation
function of the RHR Shutdown Cooling System is not credited in any
FSAR safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes are acceptable since no
automatic isolation functions are being changed. Since the manual
isolation function of the RHR Shutdown Cooling System is not
credited in any FSAR safety analysis, this change does not affect
the margin of safety assumed by the safety analysis.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179
Acting NRC Branch Chief: John G. Lamb.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 2, 2013 (TS-SQN-13-01 and 13-
02).
Description of amendment request: The proposed amendments would
revise Units 1 and 2 Technical Specifications (TSs) 3.7.5, ``Ultimate
Heat Sink,'' to place additional limitations on the maximum average
Essential Raw Cooling Water (ERCW) System supply header water
temperature during operation with one ERCW pump per loop and operation
with one ERCW supply strainer per
[[Page 74185]]
loop. In addition, the one-time limitations on Unit 1 ultimate heat
sink (UHS) temperature and the associated license condition
requirements used for the Unit 2 steam generator replacement project
are proposed to be deleted. The proposed changes would place additional
temperature limitations on the UHS TS Limiting Condition for Operation
3.7.5 with associated required actions, to support maintenance on plant
component without requiring a dual unit shutdown.
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 determination, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed change to impose additional limits on UHS
temperature while in certain ERCW system alignments does not result
in any physical changes to plant safety-related structures, systems,
or components (SSCs). The UHS and associated ERCW system function is
to remove plant system heat loads during normal and accident
conditions. As such, the UHS and ERCW system are not accident
initiators, but instead perform accident mitigation functions by
serving as the heat sink for safety-related equipment to ensure the
conditions and assumptions credited in the accident analyses are
preserved. During operation under the proposed change with only one
ERCW pump operable in a loop a single failure could cause a total
loss of ERCW flow in one loop whereas with two pumps per loop
operable only a reduction in flow would occur. In either case, one
pump or two pumps per loop operable, the other ERCW loop will
continue to perform the design function of the ERCW system.
Therefore, the proposed change does not involve a significant
increase in the probability of an accident previously evaluated.
The purpose of this change is to modify the UHS TS to be
consistent with the conditions and assumptions of the current design
basis heat transfer and flow modeling analyses for the UHS and ERCW
system. The proposed change provides assurance that the minimum
conditions necessary for the UHS and ERCW system to perform their
heat removal safety function is maintained. Accordingly, as
demonstrated by TVA design heat transfer and flow modeling
calculations, the proposed new requirements will provide the
necessary assurance that fuel cladding, Reactor Coolant System (RCS)
pressure boundary, and containment integrity limits are not
challenged during worst-case post-accident conditions. Accordingly,
the conclusions of the accident analyses will remain as previously
evaluated such that there will be no significant increase in the
post-accident dose consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve any physical changes to
plant safety related SSCs or alter the modes of plant operation in a
manner that is outside the bounds of the current UHS and ERCW system
design heat transfer and flow modeling analyses. The proposed
additional limits on UHS temperature for the specified ERCW system
alignments provide assurance that the conditions and assumptions
credited in the accident analyses are preserved. Thus, although the
specified ERCW system alignments result in reduced heat transfer
flow capability, the plant's overall ability to reject heat to the
UHS during normal operation, normal shutdown, and hypothetical
worst-case accident conditions will not be significantly affected by
this proposed change. Since the safety and design requirements
continue to be met and the integrity of the RCS pressure boundary is
not challenged, no new credible failure mechanisms, malfunctions, or
accident initiators are created, and there will be no effect on the
accident mitigating systems in a manner that would significantly
degrade the plant's response to an accident.
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 change modifies the UHS TS to maintain the UHS
temperature and associated ERCW system flows within the bounds of
the conditions and assumptions credited in the accident analyses. As
demonstrated by TVA design basis heat transfer and flow modeling
calculations, the additional limits on UHS temperature for the
specified ERCW system alignments will provide assurance that the
design limits for fuel cladding, RCS pressure boundary, and
containment integrity are not exceeded under both normal and post-
accident conditions. As required, these calculations include
evaluation of the worst-case combination of meteorology and
operational parameters, and establish adequate margins to account
for measurement and instrument uncertainties. While operating
margins have been reduced by the proposed change in order to support
necessary maintenance activities, the current limiting design basis
accidents remain applicable and the analyses conclusions remain
bounding such that the accident safety margins are maintained.
Accordingly, the proposed change will not significantly degrade the
margin of safety of any SSCs that rely on the UHS and ERCW system
for heat removal to perform their safety related functions.
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 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: Jessie F. Quichocho.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: July 30, 2013.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 4.3.1.1, ``Criticality,'' to
clarify the requirements for storage of new and spent fuel assemblies
in the spent fuel racks. This change is necessary to update the current
WBN Unit 1 TS to ensure consistency with the proposed TS 4.3.1.1 for
WBN Unit 2. In addition, editorial changes are being made to TS 4.3.1.
The proposed changes also modify the current licensing basis, as
described in Section 4.3.2.7 of the Updated Final Safety Analysis
Report (UFSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed amendment directs the operators to directly use an
existing control figure in the TS instead of a conflicting wording
of slightly lower fuel storage enrichment limit in the same section
of the TS. No change is being made to the parameters or methodology
in evaluated accidents. As a result, there is no increase in the
likelihood of existing event initiators.
This figure was supported by the original analyses that
determines the subcriticality available in the spent fuel pool and
the associated acceptable cell loading patterns have not been
changed. Thus the acceptance criteria as stated in the UFSAR are
met. Implementing the change involves no facility equipment,
procedure, or process changes that could affect the radioactive
material actually released during an event. As a result, no
conditions have been created that could
[[Page 74186]]
significantly increase the consequences of any of the events
evaluated in the UFSAR.
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 require any new or different
accidents to be postulated because no changes are being made to the
plant that would introduce any new accident causal mechanism. This
license amendment request does not affect any plant systems that are
potential accident initiators. The change in TS wording is
consistent with an existing figure in the same section of the TS
that is bounded by the original plant spent fuel pool criticality
analysis. No change to the fuel, spent fuel racks, or spent fuel
pool water chemistry are associated with this change.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment directs the operators to directly use an
existing control figure in the TS instead of a conflicting wording
of slightly lower fuel storage enrichment limit in the same section
of the TS. The change in TS wording is consistent with an existing
figure in the same section of the TS which is bounded the original
plant spent fuel pool criticality analysis. The proposed changes do
not alter the permanent plant design, including instrument set
points.
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: Jessie F. Quichocho.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: August 28, 2013.
Description of amendment request: The proposed changes would modify
WBN, Unit 1 Technical Specifications (TS) requirements related to
direct current (DC) electrical systems. In addition, a new ``Battery
Monitoring and Maintenance Program'' is being proposed. The proposed TS
changes place requirements on the battery itself rather than the
battery cells as currently required.
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 consequence of an accident previously evaluated?
Response: No.
The proposed changes restructure the Technical Specifications
(TS) for the direct current (DC) electrical power system and are
consistent with Technical Specifications Task Force (TSTF) change
TSTF-360, Revision 1 and TSTF-500, Revision 2. The proposed changes
modify TS Actions relating to battery and battery charger
inoperability. The DC electrical power system, including associated
battery chargers, is not an initiator of any accident sequence
analyzed in the Updated Final Safety Analysis Report (UFSAR).
Rather, the DC electrical power system supports equipment used to
mitigate accidents. The proposed changes to restructure TS and
change surveillances for batteries and chargers to incorporate the
updates included in TSTF-360, Revision 1 as updated by TSTF-500,
Revision 2, will maintain the same level of equipment performance
required for mitigating accidents assumed in the UFSAR. Operation in
accordance with the proposed TS would ensure that the DC electrical
power system is capable of performing its specified safety function
as described in the UFSAR. Therefore, the mitigating functions
supported by the DC electrical power system will continue to provide
the protection assumed by the analysis. The relocation of preventive
maintenance surveillances, and certain operating limits and actions,
to a licensee controlled Battery Monitoring and Maintenance Program
will not challenge the ability of the DC electrical power system to
perform its design function. Appropriate monitoring and maintenance
that are consistent with industry standards will continue to be
performed. In addition, the DC electrical power system is within the
scope of 10 CFR 50.65, ``Requirements for monitoring the
effectiveness of maintenance at nuclear power plants,'' which will
ensure the control of maintenance activities associated with the DC
electrical power system.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the UFSAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these changes.
Therefore, the proposed amendment 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 changes involve restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the UFSAR. Rather, the DC electrical power
system supports equipment used to mitigate accidents. The proposed
changes to restructure the TS and change surveillances for batteries
and chargers to incorporate the updates included in TSTF-360
Revision 1 as updated by TSTF-500, Revision 2, will maintain the
same level of equipment performance required for mitigating
accidents assumed in the UFSAR. Administrative and mechanical
controls are in place to ensure the design and operation of the DC
systems continues to meet the plant design basis describe in the
UFSAR.
Therefore, the proposed amendment will 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 margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The equipment margins will be maintained in
accordance with the plant-specific design bases as a result of the
proposed changes. The proposed changes will not adversely affect
operation of plant equipment. These changes will not result in a
change to the setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new battery Maintenance
and Monitoring Program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to
safety-related loads in accordance with analysis assumptions. TS
changes made to be consistent with the changes in TSTF-360, Revision
1, as updated by TSTF-500, Revision 2, maintain the same level of
equipment performance stated in the UFSAR and the current TSs.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Jessie F. Quichocho.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 23, 2013.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.6.5, ``CORE OPERATING LIMITS REPORT
(COLR),'' to replace WCAP-11596-P-A, ``Qualification of the Phoenix-P/
ANC Nuclear Design System for Pressurized Water Reactor Cores,'' with
WCAP-16045-P-A, ``Qualification of the Two-
[[Page 74187]]
Dimensional Transport Code PARAGON,'' and WCAP-16045-P-A, Addendum 1-A,
``Qualification of the NEXUS Nuclear Data Methodology,'' to determine
core operating limits.
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 analytical methodologies, which this license amendment
proposes for determination of core operating limits, are
improvements over the current methodologies in use at WCGS. The NRC
staff reviewed and approved these methodologies and concluded that
these analytical methods are acceptable as a replacement for the
current analytical method. Thus core operating limits determined
using the proposed analytical methods continue to assure that the
reactor operates safely and, thus, the proposed changes do not
involve an increase in the probability of an accident.
Operation of the reactor with core operating limits determined
by use of the proposed analytical methods does not increase the
reactor power level, does not increase the core fission product
inventory, and does not change any transport assumptions. Therefore
the proposed methodology and TS changes do not involve a significant
increase in the consequences of an accident.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides revised analytical methods for
determining core operating limits, and does not change any system
functions or maintenance activities. The change does not involve
physical alteration of the plant, that is, no new or different type
of equipment will be installed. The change does not alter
assumptions made in the safety analyses but ensure that the core
will operate within safe limits. This change does not create new
failure modes or mechanisms that are not identifiable during
testing, and no new accident precursors are generated.
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. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes do not physically alter safety-
related systems, nor does it affect the way in which safety related
systems perform their functions. The setpoints at which protective
actions are initiated are not altered by the proposed changes.
Therefore, sufficient equipment remains available to actuate upon
demand for the purpose of mitigating an analyzed event. The proposed
analytical methodology is an improvement that allows more accurate
modeling of core performance. The NRC has reviewed and approved this
methodology for use in lieu of the current methodology; thus, the
margin of safety is not reduced due to this change.
Therefore, it is concluded that 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined 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.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, 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, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available documents created or received at the
NRC are accessible electronically through the Agencywide Documents
Access and Management System (ADAMS) in the NRC Library at 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's Reference staff at 1-800-397-4209, 301-415-4737 or by
email 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 application for amendment: December 12, 2012.
Brief description of amendment: The amendments revised the
Technical Specifications (TSs) relating to reactor coolant system (RCS)
activity limits by replacing the current TS limits on primary coolant
gross specific activity with limits on primary coolant noble gas
activity. The noble gas activity would reflect a new DOSE EQUIVALENT
XE-133 definition that would replace the current E-bar average
disintegration energy definition. The changes are consistent with NRC-
approved Industry/Technical Specifications Task Force (TSTF) Standard
Technical Specification change traveler, TSTF-490, Revision 0,
``Deletion of E-bar Definition and Revision to RCS [Reactor Coolant
System] Specific Activity Technical Specifications,'' with deviations.
Date of issuance: November 25, 2013.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment No.: Unit 1-192; Unit 2-192; Unit 3-192.
Renewed Facility Operating License Nos. NPF-41, NPF-51; and NPF-74:
The
[[Page 74188]]
amendment revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 4, 2013 (78 FR
14128).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 25, 2013.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit 2, New London County, Connecticut
Date of amendment request: April 3, 2013.
Description of amendment request: The amendment would revise
Technical Specification 3.9.16 ``Shielded Cask,'' due to changes to the
minimum decay time for fuel assemblies adjacent to the spent fuel pool
cask laydown area.
Date of issuance: November 14, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 316.
Renewed Facility Operating License No. DPR-65: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: June 11, 2013 (78 FR
35062).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 14, 2013.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Units 1 and 2, Salem County, New Jersey
Date of amendment requests: November 30, 2012, as supplemented by
letter dated May 31, 2013.
Brief description of amendments: The amendments approve a change to
the site Emergency Plan to remove the backup plant vent extended range
noble gas radiation monitoring (R45) indication, recording, and alarm
capability in the emergency response facilities. Although the R45B/C
monitor equipment skid will be removed, the licensee will maintain a
capability in its Emergency Plan to take post-accident samples from the
plant vent stack, as specified by an earlier commitment to Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.''
Date of issuance: November 27, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 305 and 287.
Renewed Facility Operating License Nos. DPR-70 and DPR-75: The
amendments revised the Facility Operating License and approved
revisions to the Emergency Plan.
Date of initial notice in Federal Register: May 14, 2013 (78 FR
28252). The supplemental letter dated May 31, 2013, provided
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 November 27, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 2nd day of December 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-29168 Filed 12-9-13; 8:45 am]
BILLING CODE 7590-01-P