[Federal Register Volume 80, Number 217 (Tuesday, November 10, 2015)]
[Notices]
[Pages 69707-69719]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-28347]
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NUCLEAR REGULATORY COMMISSION
[NRC-2015-0253]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: 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 October 10, 2015, to October 26, 2015. The
last biweekly notice was published on October 27, 2015.
DATES: Comments must be filed December 10, 2015. A request for a
hearing must be filed by January 11, 2016.
ADDRESSES: You may submit comments 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-2015-0253. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
[[Page 69708]]
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-2549, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2015-0253 when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0253.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection 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 (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section of this document.
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-2015-0253, facility name, unit
number(s), application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. 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.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 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, (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.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
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's 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 within 60 days, the Commission or a presiding officer designated
by the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
[[Page 69709]]
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 with respect to resolution of that person's admitted
contentions, including the opportunity to present evidence and to
submit a cross-examination plan for cross-examination of witnesses,
consistent with NRC regulations, policies and procedures.
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).
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of any amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, federally-recognized Indian
tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by
December 28, 2015. The petition must be filed in accordance with the
filing instructions in the ``Electronic Submissions (E-Filing)''
section of this document, and should meet the requirements for
petitions for leave to intervene set forth in this section, except that
under Sec. 2.309(h)(2) a State, local governmental body, or Federally-
recognized Indian tribe, or agency thereof does not need to address the
standing requirements in 10 CFR 2.309(d) if the facility is located
within its boundaries. A State, local governmental body, Federally-
recognized Indian tribe, or agency thereof may also have the
opportunity to participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not
qualified, to become a party to the proceeding may, in the discretion
of the presiding officer, be permitted to make a limited appearance
pursuant to the provisions of 10 CFR 2.315(a). A person making a
limited appearance may make an oral or written statement of position on
the issues, but may not otherwise participate in the proceeding. A
limited appearance may be made at any session of the hearing or at any
prehearing conference, subject to the limits and conditions as may be
imposed by the presiding officer. Persons desiring to make a limited
appearance are requested to inform the Secretary of the Commission by
December 28, 2015.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC'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/getting-started.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.
[[Page 69710]]
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC'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 NRC'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 public
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, in some instances, 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 these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: July 17, 2015. A publicly-available
version is in ADAMS under Accession No. ML15232A017.
Description of amendment request: The proposed amendment corrects a
usage problem with recently issued Amendment Nos. 382, 384, and 383
(ADAMS Accession No. ML13231A013), which precludes Oconee Nuclear
Station Technical Specification (TS) 3.8.1, ``AC [Alternating Current]
Sources-Operating,'' Condition H from being used as planned. The
proposed change revises the note to TS 3.8.1 Required Actions L.1, L.2,
and L.3, to remove the 12-hour time limitation when the second Keowee
Hydroelectric Unit (KHU) is made inoperable for the purpose of
restoring the KHU undergoing maintenance to OPERABLE status. Removal of
the 12-hour time limitation allows use of the full 60-hour Completion
Time of Required Action H.2 when the unit(s) have been in Condition C
for greater than 72 hours and both units are made inoperable for
purposes of restoring the KHU undergoing maintenance to OPERABLE
status.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment revises the note to Technical
Specification (TS) 3.8.1 Required Actions L.1, L.2, and L.3 to
indicate the Required Actions are not required when the Condition is
entered to restore a KHU to OPERABLE status. This change is
consistent with Amendment Nos. 382, 384, and 383, which approved a
cumulative 240 hours of allowed outage time over a 3-year period
when both KHUs are inoperable when in the 45-day Completion Time of
TS 3.8.1 Required Action C.2.2.5. The proposed TS change does not
modify the reactor coolant system pressure boundary, nor make any
physical changes to the facility design, material, or construction
standards. The probability of any design basis accident (DBA) is not
affected by this change, nor are the consequences of any DBA
affected by this change. The proposed change does not involve
changes to any structures, systems, or components (SSCs) that can
alter the probability for initiating a LOCA [loss-of-coolant
accident] event.
[[Page 69711]]
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed TS change revises the note to TS 3.8.1 Required
Actions L.1, L.2, and L.3 to indicate the Required Actions are not
required when the Condition is entered to restore a KHU to OPERABLE
status. Revision of the note allows the 60 hour Completion Time of
TS 3.8.1 Condition H to limit the time that both KHUs are
inoperable. The changes do not alter the plant configuration (no new
or different type of equipment will be installed) or make changes in
methods governing normal plant operation. No new failure modes are
identified, nor are any SSCs required to be operated outside the
design bases.
Therefore, the possibility of a new or different kind of
accident from any kind of accident previously evaluated is not
created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed TS change revises the note to TS 3.8.1 Required
Actions L.1, L.2, and L.3 to indicate the Required Actions are not
required when the Condition is entered to restore a KHU to OPERABLE
status. Revision of the note allows the 60 hour Completion Time of
TS 3.8.1 Condition H to limit the time that both KHUs are
inoperable. The proposed TS change does not involve: (1) A physical
alteration of the Oconee Units; (2) the installation of new or
different equipment; (3) operating any installed equipment in a new
or different manner; (4) a change to any set points for parameters
which initiate protective or mitigation action; or (5) any impact on
the fission product barriers or safety limits.
Therefore, this request 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--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: August 27, 2015. A publicly-available
version is in ADAMS under Accession No. ML15246A231.
Description of amendment request: The amendment would approve
changes to the Permanently Defueled Emergency Plan (PDEP) to reflect
the planned use of an Independent Spent Fuel Storage Installation
(ISFSI) located in the Crystal River Unit 3 Nuclear Plant Protected
Area while the spent fuel pool contains spent fuel assemblies.
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 site PDEP and PD EAL [Permanently Defueled
Emergency Action Level] Bases Manual revisions are commensurate with
the ongoing and anticipated reduction in radiological source term at
the CR-3 site and reflects the addition of spent fuel being
transferred to the ISFSI facility. These changes add the
responsibility for responding to ISFSI emergencies to the CR-3 PDEP
Shift Supervisor/Certified Fuel Handler, and accompanying changes to
the PD EAL Bases Manual due to the creation of a potential or actual
release path to the environment, degradation of one or more storage
canisters or fuel assemblies due to environmental factors, and
configuration changes that could cause challenges in removing the
canister or fuel from storage.
There are no longer design basis accidents or postulated beyond
design basis accidents that could result in doses to the public and
the environment beyond the exclusion area boundary that would exceed
the EPA PAGs [Protective Action Guidelines]. CR-3 was shut down on
September 26, 2009, and will not be restarted. With the reactor
permanently defueled, the spent fuel pool and its support systems
are dedicated to spent fuel storage only. With the spent fuel in wet
storage for some time, the spectrum of postulated accidents is much
smaller than for an operational plant, with the majority of design
basis accidents no longer possible. The only remaining credible
design basis accident is the fuel handling accident, which does not
result in exceeding the EPA Protective Action Guidelines at the
exclusion area boundary. Spent fuel located in the spent fuel pools
will be transferred to the ISFSI facility. Emergency Planning Zones
beyond the exclusion area boundary and the associated protective
actions are no longer required. No corporate personnel, personnel
involved in off-site dose projections, or personnel with special
qualifications are required to augment the ERO [Emergency Response
Organization].
The credible events for the ISFSI facility remain unchanged. The
indications of damage to a loaded Dry Shielded Canister CONFINEMENT
BOUNDARY have been revised to be twice the design basis dose rate as
described in Draft Amendment 14 to COC [Certificate of Compliance]
1004 Technical Specifications for the Standardized NUHOMS Horizontal
Modular Storage System, Sections 5.2.4 `Radiation Protection
Program' and 5.4.2 HSM [horizontal storage module] or HSM-H Dose
Rate Evaluation Program (Reference 7), while in transit or HSM
storage.
Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as
indicated by the following on-contact radiation readings at some
prescribed distance from the transfer cask or HSM:
1300 mrem/hr (gamma + neutron) on the radial surface of the fuel
transfer cask while in transit to the ISFSI HSM
OR
1050 mrem/hr (gamma + neutron)--HSM Front Bird Screen
4 mrem/hr (gamma + neutron)--HSM Outside Door
40 mrem/hr (gamma + neutron)--HSM End Shield Wall Exterior while in
HSM storage.
This change is consistent with industry practices previously
approved by the NRC to distinguish whether a degraded containment
barrier condition exists.
The probability of occurrence of previously evaluated accidents
is not increased, since most previously analyzed accidents can no
longer occur and the probability of the remaining credible design
basis accident is unaffected by the proposed amendment.
The deletion of the Communicator position does not impact
Emergency Notifications from the plant since the Emergency
Coordinator has shown the capability to perform this function. This
function is not involved in operations or evolutions that could
cause an accident since it is not performed until after the
emergency is declared, and has no effect on accident mitigation.
Therefore, the proposed changes do not affect any plant system,
the operation and maintenance of CR-3 and the ISFSI facility, or
increase the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes have no impact on facility structures,
systems, or components (SSCs) affecting the safe storage of
irradiated fuel, or on the methods of operation of such SSCs, or on
the handling and storage of irradiated fuel itself. Additionally,
the proposed changes have no impact on a Fuel Handling Accident,
which is the remaining credible design basis accident evaluated. The
CR-3 PDEP is applicable for the plant's defueled condition. There is
no impact on the prevention, diagnosis, or mitigation of reactor-
related transients as there are no longer any reactor-related
accidents. Accidents cannot result in different or more adverse
failure modes or accidents than previously evaluated because the
reactor is permanently shut down and defueled, and CR-3 is no longer
authorized to operate the reactor.
There are no longer credible events that would result in doses
to the public beyond the exclusion area boundary that would exceed
the EPA [Environmental Protection
[[Page 69712]]
Agency] PAGs. Spent fuel waste will be transferred to the ISFSI
facility. Emergency Planning Zones beyond the site boundary and the
associated protective actions are no longer required. No corporate
personnel, personnel involved in offsite dose projections, or
personnel with special qualifications are required to augment the
ERO.
The credible events for the ISFSI facility remain unchanged. The
indications of damage to a loaded Dry Shielded Canister CONFINEMENT
BOUNDARY have been revised to be twice the design basis dose rate as
described in Draft Amendment 14 to COC 1004 Technical Specifications
for the Standardized NUHOMS Horizontal Modular Storage System,
Sections 5.2.4 `Radiation Protection Program' and 5.4.2 HSM or HSM-H
Dose Rate Evaluation Program (Reference 7), while in transit or HSM
storage.
Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as
indicated by the following on-contact radiation readings at some
prescribed distance from the transfer cask or HSM:
1300 mrem/hr (gamma + neutron) on the radial surface of the fuel
transfer cask while in transit to the ISFSI horizontal storage
module (HSM)
OR
1050 mrem/hr (gamma + neutron)--HSM Front Bird Screen
4 mrem/hr (gamma + neutron)--HSM Outside Door
40 mrem/hr (gamma + neutron)--HSM End Shield Wall Exterior while in
HSM storage.
This change is consistent with industry practices previously
approved by the NRC to distinguish whether a degraded containment
barrier condition exists. The proposed amendment does not introduce
a new mode of plant operation or new accident pre-cursors, does not
involve any physical alterations to plant configurations, or make
changes to plant system set points that initiate a new or different
kind of accident.
The deletion of the Communicator position does not impact
Emergency Notifications from the plant since the Emergency
Coordinator has shown the capability to perform this function. This
function is not involved in operations or evolutions that could
cause or create new or different kinds of accidents since the
communication of Emergency Notifications is not performed until
after the emergency is declared and cannot affect an accident or
event already in progress.
Therefore, the proposed changes have no direct effect on any
plant system, the operation and maintenance of CR-3 or the ISFSI
facility, or create the possibility of a new or different kind of
accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes have no direct effect on any plant system,
do not involve any physical plant limit or parameter, License
Condition, Technical Specification Limiting Condition of Operability
or operating philosophy, and therefore cannot affect any margin of
safety. The margin of safety is maintained by conforming to the CR-3
Technical Specifications or the ISFSI Technical Specifications. The
proposed CR-3 PDEP and PD EAL Bases Manual revisions are
commensurate with the on-going and anticipated reduction in
radiological source term at the CR-3 site and reflect spent fuel
being transferred to the ISFSI facility. These changes add the
responsibility for implementing the emergency plan for the ISFSI
facility to the Shift Supervisor/Certified Fuel Handler.
The only remaining credible accident for CR-3, while the SFP is
operable and prior to the transference of all spent fuel to dry
shielded canisters, is a fuel handling accident. The proposed
amendment does not adversely affect the inputs or assumptions of any
design basis analysis that impact the fuel handling accident. There
are no longer credible events that would result in doses to the
public beyond the exclusion area boundary that would exceed the EPA
PAGs. Emergency Planning Zones beyond the exclusion area boundary
and the associated protective actions are no longer required. No
corporate personnel, personnel involved in offsite dose projections,
or personnel with special qualifications are required to augment the
ERO. The credible events for the ISFSI facility remain unchanged.
The indications of damage to a loaded Dry Shielded Canister
CONFINEMENT BOUNDARY have been revised to be twice the design basis
dose rate as described in Draft Amendment 14 to COC 1004 Technical
Specifications for the Standardized NUHOMS Horizontal Modular
Storage System, Sections 5.2.4 `Radiation Protection Program' and
5.4.2 HSM or HSM-H Dose Rate Evaluation Program (Reference 7), while
in transit or HSM storage.
Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as
indicated by the following on-contact radiation readings at some
prescribed distance from the transfer cask or HSM:
1300 mrem/hr (gamma + neutron) on the radial surface of the fuel
transfer cask while in transit to the ISFSI HSM
OR
1050 mrem/hr (gamma + neutron)--HSM Front Bird Screen
4 mrem/hr (gamma + neutron)--HSM Outside Door
40 mrem/hr (gamma + neutron)--HSM End Shield Wall Exterior while in
HSM storage.
This change is consistent with industry practices previously
approved by the NRC to distinguish whether a degraded containment
barrier condition exists. The proposed changes are limited to the
CR-3 PDEP and PD EAL Bases Manual and do not impact the safe storage
of irradiated fuel. The proposed revisions do not affect any
requirements for SSCs credited in the remaining analyses of
applicable postulated accidents, and as such, do not affect the
margin of safety associated with these accident analyses.
The deletion of the Communicator position does not impact
Emergency Notifications from the plant since the Emergency
Coordinator has shown the capability to perform this function. This
function is not involved in design basis analyses or operations that
could cause any decrease in any previously analyzed safety margin.
Therefore, the proposed changes do not create the possibility of
reduction in any safety margin.
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: Lara S. Nichols, 550 South Tryon Street,
Charlotte NC 28202.
NRC Branch Chief: Bruce A. Watson, CHP.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 8, 2015. A publicly-available
version is in ADAMS under Accession No. ML15258A185.
Description of amendment request: The proposed amendment would
replace the Technical Specification (TS) Figure 4.1-1, ``Site and
Exclusion Area Boundaries and Low Population Zone,'' with a text
description in TS 4.1, ``Site Location.'' In addition, a typographical
error would be corrected from ``LGHR'' to ``LHGR'' [Linear Heat
Generation Rate] in TS 1.1, ``Definitions.''
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 removes a figure, replaces that figure with
a text description of the site location and corrects a typographical
error. An administrative change such as this is not an initiator of
any accident previously evaluated. As a result, the probability of
an accident previously evaluated is not affected. The consequences
of an accident with the incorporation of this administrative change
are not different than the consequences of the same accident without
this change. As a result, the consequences of an accident previously
evaluated are not affected by this change.
Based on the above, it is concluded that the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not modify the plant design, nor does
the proposed change alter the operation of the plant or equipment
involved in either routine plant operation or
[[Page 69713]]
in the mitigation of design basis accidents. The proposed change is
administrative only.
Based on the above, it is concluded that 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 consists of an administrative change to
remove a figure, replace that figure with a text description of the
site location, and correct a typographical error. The change does
not alter the manner in which safety limits, limiting safety system
settings, or limiting conditions for operation are determined. The
safety analysis acceptance criteria are not affected by this change.
The proposed change will not result in plant operation in a
configuration outside of the design basis.
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: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: July 24, 2015. A publicly-available
version is in ADAMS under Accession No. ML15246A408.
Description of amendment request: The amendment would make
editorial corrections to Technical Specification (TS) Section 1.4,
``Frequency.'' Example 1.4-1 would be revised to be consistent with
NRC-approved Industry Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-485, Revision 0,
``Correct Example 1.4-1.'' In addition, Example 1.4-5 and Example 1.4-6
would be revised to correct typographical errors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are editorial in nature and have no effect
on accident scenarios previously evaluated. The proposed changes
consist of editorial corrections to TS Section 1.4, ``Frequency,''
that would make the Duane Arnold Energy Center (DAEC) TS consistent
with the Standard Technical Specifications for General Electric BWR/
4 Plants (NUREG-1433). The proposed changes do not affect initiating
events for accidents previously evaluated and do not affect or
modify plant systems or procedures used to mitigate the progression
or outcome of those accident scenarios.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are editorial in nature consisting of
editorial corrections to TS Section 1.4, ``Frequency.'' The proposed
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. No new accident
scenarios, failure mechanisms, or limiting single failures are
introduced as a result of the proposed changes.
The proposed changes do not introduce any new accident
precursors, nor do they impose any new or different requirements or
eliminate any existing requirements. The proposed changes do not
alter assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the
fission product barriers (fuel cladding, reactor coolant system, and
primary containment) to perform their design functions during and
following postulated accidents. The proposed changes are editorial
in nature consisting of editorial corrections to TS Section 1.4,
``Frequency.'' No setpoints at which protective actions are
initiated are altered by the proposed changes. The proposed changes
do not alter the manner in which the safety limits are determined.
These changes are consistent with plant design and do not change the
TS operability requirements; thus, previously evaluated accidents
are not affected by this proposed change.
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: Mr. William Blair, P.O. Box 14000, Juno
Beach, FL 33408-0420.
NRC Branch Chief: David L. Pelton.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: August 6, 2015. A publicly-available
version is in ADAMS under Accession No. ML15246A410.
Description of amendment request: The proposed amendment would
resolve a 10 CFR part 21 condition concerning a potential to
momentarily violate Reactor Core Safety Limit 2.1.1.1 during Pressure
Regulator Failure Maximum Demand (Open) transient.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the reactor steam dome pressure from 785
psig to 685 psig in TS [Technical Specification] SLs [Safety Limits]
2.1.1.1 and 2.1.1.2 does not alter the use of the analytical methods
used to determine the safety limits that have been previously
reviewed and approved by the NRC. The proposed change is in
accordance with an NRC approved critical power correlation
methodology and as such maintains required safety margins. The
proposed change does not adversely affect accident initiators or
precursors nor does it alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained.
The proposed change does not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits. The proposed change does
not require any physical change to any plant SSCs nor does it
require any change in systems or plant operations. The proposed
change is consistent with the safety analysis assumptions and
resultant consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (i.e., no new or different type of equipment will be
[[Page 69714]]
installed) or a change in the methods governing normal plant
operation. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
change.
The proposed change does not introduce any new accident
precursors, nor does it impose any new or different requirements or
eliminate any existing requirements. The proposed change does not
alter assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the
fission product barriers (fuel cladding, reactor coolant system, and
primary containment) to perform their design functions during and
following postulated accidents. Evaluation of the 10 CFR part 21
condition by General Electric determined that there was no decrease
in the safety margin, the Minimum Critical Power Ratio improves
during the transient, and therefore is not a threat to fuel cladding
integrity.
The proposed change to Reactor Core Safety Limits 2.1.1.1 and
2.1.1.2 is consistent with, and within the capabilities of the
applicable NRC approved critical power correlation, and thus
continues to ensure that valid critical power calculations are
performed. No setpoints at which protective actions are initiated
are altered by the proposed change. The proposed change does not
alter the manner in which the safety limits are determined. This
change is consistent with plant design and does not change the TS
operability requirements; thus, previously evaluated accidents are
not affected by this proposed change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William Blair, P.O. Box 14000, Juno
Beach, FL 33408-0420.
NRC Branch Chief: David L. Pelton.
NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: June 12, 2015, as supplemented by
letters dated August 11, 2015, and August 28, 2015. Publicly-available
versions are in ADAMS under Accession Nos. ML15166A042, ML15223B277,
and ML15240A017, respectively.
Description of amendment request: The amendments would revise the
Point Beach Emergency Plan, to increase the staff augmentation times
for Emergency Response Organization (ERO) response functions, from 30
and 60 minutes, to 60 minutes and 90 minutes, respectively. Additional
changes include relocation of the Emergency Director and Emergency
Action Level Monitor positions, and the addition of an Assistant
Emergency Operations Facility Manager position.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed increase in staff augmentation times has no effect
on normal plant operation or on any accident initiator or precursors
and does not impact the function of plant structures, systems, or
components (SCCs). The proposed change does not alter or prevent the
ability of the ERO to perform their intended functions to mitigate
the consequences of an accident or event. The ability of the ERO to
respond adequately to radiological emergencies has been demonstrated
as acceptable through a staffing analysis as required by 10 CFR 50
Appendix E.IV.A.9.
Therefore, the proposed Emergency Plan changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not impact the accident analysis. The
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed), a change in
the method of plant operation, or new operator actions. The proposed
change does not introduce failure modes that could result in a new
accident, and the change does not alter assumptions made in the
safety analysis. This proposed change increases the staff
augmentation response times in the Emergency Plan, which are
demonstrated as acceptable through a staffing analysis as required
by 10 CFR 50 Appendix E.IV.A.9. The proposed change does not alter
or prevent the ability of the ERO to perform their intended
functions to mitigate the consequences of an accident or event.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change is
associated with the Emergency Plan staffing and does not impact
operation of the plant or its response to transients or accidents.
The change does not affect the Technical Specifications. The
proposed change does not involve a change in the method of plant
operation, and no accident analyses will be affected by the proposed
change. Safety analysis acceptance criteria are not affected by this
proposed change. The revised Emergency Plan will continue to provide
the necessary response staff with the proposed change. A staffing
analysis and a functional analysis were performed for the proposed
change on the timeliness of performing major tasks for the
functional areas of Emergency Plan. The analysis concluded that an
extension in staff augmentation times would not significantly affect
the ability to perform the required Emergency Plan tasks. Therefore,
the proposed change is determined to not adversely affect the
ability to meet 10 CFR 50.54(q)(2), the requirements of 10 CFR 50
Appendix E, and the emergency planning standards as described in 10
CFR 50.47(b).
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: William Blair, Managing Attorney--Nuclear,
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard,
Juno Beach, FL 33408-0420.
NRC Branch Chief: David L. Pelton.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo
County, California
Date of amendment request: September 16, 2015. A publicly-available
version is in ADAMS under Accession No. ML15259A576.
Description of amendment request: The amendment would revise the
Reactor Coolant System (RCS) minimum flow specified in Technical
Specification (TS) 3.4.1, ``RCS Pressure, Temperature, and Flow
Departure from Nucleate Boiling (DNB) Limits.'' The proposed change is
necessary to correct a non-conservative TS value for DCPP, Unit 1. The
Unit 1 RCS flow specified in TS 3.4.1 for 100 percent power is 359,000
gallons per minute (gpm). However, the TS value is less than the
359,200 gpm RCS minimum measured flow (MMF) value specified in the
Updated Final Safety Analyses Report
[[Page 69715]]
(UFSAR) Table 4.1-1, ``Reactor Design Comparison.'' The UFSAR RCS MMF
value represents the RCS flow value used in the reactor core DNB safety
analyses. This issue has been entered in the DCPP corrective action
program, and the actual Unit 1 RCS flow value has been verified to be
within the limits required by the applicable safety analyses.
In order to resolve the non-conservative TS value, the proposed
change would revise the RCS flow requirements in DCPP TS 3.4.1 to be
consistent with TS 3.4.1 in NUREG-1431, Revision 4, Volume 1,
``Standard Technical Specifications--Westinghouse Plants,'' April 2012
(ADAMS Accession No. ML12100A222). The proposed change to the RCS flow
requirements in TS 3.4.1 would also be consistent with the NRC-approved
Technical Specification Task Force (TSTF) Traveler-339-A, Revision 2,
``Relocate TS Parameters to [Core Operating Limits Report] COLR,'' and
NRC-approved WCAP-14483-A, ``Generic Methodology for Expanded Core
Operating Limits Report,'' dated June 13, 2000 (ADAMS Accession No.
ML003723269).
The proposed change would delete the current DCPP, Units 1 and 2 TS
3.4.1 RCS flow Tables 3.4.1-1 and 3.4.1-2, and would add the DCPP,
Units 1 and 2 RCS thermal design flow values of 350,800 gpm and 354,000
gpm, respectively, to the requirements of TS 3.4.1. In addition, the
proposed change would add the RCS MMF values of 359,200 gpm and 362,500
gpm, to the DCPP, Units 1 and 2 COLR, respectively. Consistent with the
Standard Technical Specifications (STS), the proposed change would also
include a reference to the RCS COLR flow requirements in the TS 3.4.1
Limiting Condition for Operation and Surveillance Requirements. Due to
the elimination of RCS flow Tables 3.4.1-1 and 3.4.1-2, a reference to
these tables is also deleted from Figure 2.1.1-1, ``Reactor Core Safety
Limit.''
As such, the proposed change would resolve the non-conservative TS
value for Unit 1 and serve to make the DCPP, Units 1 and 2 TS more
consistent with the STS in NUREG-1431.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the DCPP Unit 1 and Unit 2 RCS flow
requirements in TS 3.4.1, ``RCS Pressure, Temperature, and Flow
Departure from Nucleate Boiling (DNB) Limits,'' to be more
consistent with TS 3.4.1 in NUREG-1431 and with the applicable DCPP
safety analyses. The proposed RCS flow values will ensure the
assumptions of the safety analyses continue to be met.
As such, the proposed change does not affect the design or
function of any plant structures, systems, and components (SSCs).
Thus, the proposed change does not affect plant operation, design
features, or any analysis that verifies the capability of an SSC to
perform a design function. As the proposed change is consistent with
the RCS flow assumptions of the safety analyses, the proposed change
does not affect any previously evaluated accidents in the UFSAR. In
addition, the proposed change does not affect any SSCs, operating
procedures, and administrative controls which have the function of
preventing or mitigating any accident previously evaluated in the
UFSAR.
The proposed change will not alter any accident analyses
assumptions discussed in the UFSAR and will continue to assure the
DCPP units operate within the assumptions of the applicable safety
analyses described in the UFSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The proposed change revises the DCPP Unit 1 and Unit 2 RCS flow
requirements in TS 3.4.1, ``RCS Pressure, Temperature, and Flow
Departure from Nucleate Boiling (DNB) Limits,'' to be more
consistent with TS 3.4.1 in NUREG-1431 and with the applicable DCPP
safety analyses. The proposed RCS flow values will ensure the
assumptions of the safety analyses continue to be met.
The proposed change 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 proposed change involves no
physical plant modification or changes in plant operation, therefore
no new failure modes are created. As such, the proposed change does
not create new failure modes or mechanisms that are not identifiable
during testing, and no new accident precursors are generated.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change does 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 change.
Therefore, sufficient equipment remains available to actuate upon
demand for the purpose of mitigating an analyzed event. The proposed
RCS flow value changes are consistent with the plant safety
analyses. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, CA 94120.
NRC Branch Chief: Michael T. Markley.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San
Diego County, California
Date of amendment request: August 20, 2015. A publicly-available
version is in ADAMS under Accession No. ML15236A018.
Description of amendment request: The proposed amendment would
revise Appendix 3A of the Updated Final Safety Analysis Report to more
fully reflect the permanently shutdown status of the SONGS, Units 2 and
3. The revision would include a limited set of exceptions and
clarifications to referenced Regulatory Guides to reflect the
significantly reduced decay heat loads in the SONGS, Units 2 and 3
Spent Fuel Pools and to support corresponding design basis changes and
modifications that will allow for the implementation of the ``cold and
dark'' strategy outlined in the SONGS Post-Shutdown Decommissioning
Activities Report (PSDAR).
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?
The only accident previously evaluated, is the Spent Fuel Pool
Boiling Event. The initiating event (loss of cooling) would no
longer lead to a rapid increase in pool temperature to the boiling
point or to a relatively short-term reduction in pool level due to
evaporative losses. Currently a loss of
[[Page 69716]]
cooling would lead to a very slow heat-up toward the boiling point
taking at least a week or more. From that point the slower
evaporative losses would take several weeks to reduce inventory to
unacceptable levels.
The most likely cause of a loss of function of the Spent Fuel
Pool Cooling System (SFPCS) is not a failure of components in the
cooling system, but instead a loss of electrical power. The
probability of a loss of power is substantially higher than the
probability of a contemporaneous common cause failure of active
components in the cooling system. For example, NRC has collected
operating experience on loss of Spent Fuel Pool (SFP) cooling for
nuclear plants in the U.S., which includes both safety-related and
non-safety-related cooling systems. As indicated in NUREG-1275,
Volume 12, the causes of loss of SFP cooling were the loss of the
SFP cooling pumps due to loss of electrical power (39 of 56 events),
loss of suction from the spent fuel pool, flow blockage, loss of the
heat sink, and one case of inadequate configuration control. As
concluded by the NRC: ``The dominant cause of the actual loss of SFP
cooling events was loss of electrical power to the SFP cooling
pumps.'' There were no cases involving a common cause failure mode,
such as seismic events or tornados. Given this operating experience,
any increase in the probability of a spent fuel pool boiling event
due to the seismic re-classification of the system would be minimal
in comparison to the failure rate due to loss of electrical power.
The change in commitment does not affect the consequences of the
spent fuel pool boiling accident (which by definition assumes loss
of the spent fuel pool cooling system). Revised dose calculations
were completed to support the changes to the Updated Final Safety
Analysis Report (UFSAR) Chapter 15 Accident Analysis, and the UFSAR
was revised to reflect the new analysis. These were recently
reviewed to verify they remain bounding for the much slower event,
even if it is not terminated (through restored cooling or adequate
make-up) prior to reaching levels approaching the top of the stored
fuel. This re-evaluation confirmed the doses previously calculated
remain bounding and several orders of magnitude below applicable
limits.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The only accident relevant to this proposed change would be an
unmitigated Spent Fuel Pool Boiling Event (i.e., boiling without
restoration of cooling or make-up prior to uncovering of the spent
fuel). The initiating event (loss of cooling) would no longer lead
to a rapid increase in pool temperature to the boiling point and a
relatively short-term reduction in pool level due to evaporative
losses. Currently a loss of cooling would lead to a very slow heatup
toward the boiling point taking at least a week or more. From that
point the slower evaporative losses would take several weeks to
reduce inventory to unacceptable levels. The only safety function
remaining relates to maintaining the fuel cladding in the SFP
(cooling is not a safety-related function as defined in the updated
Chapter 15 Fuel Pool Boiling Accident Analysis, only maintaining
water level--Reference 6.12). The only remaining safety related SSCs
at SONGS Units 2 and 3 are the Spent Fuel Pool and related
structural components (pool liner, structure, and racks).
The Make-up System will ensure that sufficient water is supplied
to the SFPs in the event of loss of cooling. In addition to the
Seismic Category I make-up source, currently there are numerous
other diverse sources of make-up for the SFPs, including:
As provided in SONGS Units 2 and 3 procedures, the
Nuclear Service Water connections located on the SFP operating level
can be used via hoses to fill the pool. These connections are QC
III, Seismic Category II.
As provided in SONGS Units 2 and 3 Mitigation
Strategies, water from Fire Water Tanks T-102 and T-103 via Fire
Pumps P-220 (diesel driven), P-221 or P-222 (both of which are motor
driven) can be provided through the installed fire system piping to
two fire hose cabinets located on the Spent Fuel Pool Operating
level. The tanks, pumps and piping are QC III-EPS and Seismic
Category II.
As provided in SONGS Units 2 and 3 Mitigation
Strategies, make-up to the SFPs can be provided using water from one
or more of the following sources: Demineralized Water Tanks T-266,
T-267 or T-268, all are located at a higher elevation at the Make-up
Demineralizer Area at the south end of the plant. Skid mounted pump
P-i1058 delivers water from these sources to the seismic standpipe
and from the standpipe to the SFP. T-266, T-267 and T-268 are QC
III, Seismic Category II. P-1058 is QC III-EPS and Seismic Category
III.
As discussed in SONGS Units 2 and 3 Mitigation
Strategies, the 10'' City Water Line Supply Line can be used as an
alternate source of SFP make-up water.
Another make-up path is available using the Seismic
Category I Demineralized Water Storage Tank (T-351) located in the
North Industrial Area along with Seismic Category I portable diesel
driven Fire Pump (P-i1065) using strategically staged hoses between
the tank, pump, Seismic Category I standpipe and the Spent Fuel
Pool. The hoses are pressure tested annually and are inspected for
location quarterly per SONGS Units 2 and 3 procedures.
The Mitigation Strategies are sequenced to assure the strategies
can be deployed in 2 hours or less. The capability to achieve this
time requirement was evaluated in a formal study and further
demonstrated in the field using actual staff, procedures and
equipment.
Given the number and diversity of make-up sources, and the time
available to supply make-up to the SFPs in the loss of spent fuel
pool cooling, it is not credible to postulate a complete loss of
make-up to a SFP. As discussed in NRC's June 30, 2014, letter
concerning San Onofre Nuclear Generating Station, Units 2 and 3--
Rescission of Order EA-12-049:
[T]he time to boil off water inventory in the SFP to a level of
10 feet above the spent fuel will be sufficiently long to obviate
the need for additional strategies to restore SFP cooling. The NRC
staff concludes that given the low decay heat levels and the long
time to boil off, the reliance on the SFP inventory for passive
cooling provides an equivalent level of protection as that which
would be provided by the initial phase of the guidance and
strategies for maintaining or restoring SFP cooling capabilities
that would be necessary for compliance with Order EA-12-049 using
installed equipment. The staff further concludes that the long time
to boil off the SFP inventory to a point at which make-up would be
necessary for radiation shielding purposes obviates the need for the
transition phase of the guidance and strategies that would be
necessary for compliance with Order EA-12-049 using on-site portable
equipment. The staff also concludes that the low decay heat and long
boil-off period provides sufficient time for the licensee to obtain
off-site resources on an ad hoc basis to sustain the SFP cooling
function indefinitely, obviating the need for the final phase of the
guidance and strategies that would be necessary for compliance with
Order EA-12-049.
Similarly, as described in NRC's 2015 exemption from certain
emergency planning requirements for SONGS Units 2 and 3:
Additionally, in its letters to the NRC dated October 6, 2014,
and December 15, 2014, SCE described the SFP make-up strategies that
could be used in the event of a catastrophic loss of SFP inventory.
The multiple strategies for providing make-up water to the SFP
include: Using existing plant systems for inventory make-up; an
internal strategy that relies on installed fire water pumps and
service water or fire water storage tanks; or an external strategy
that uses portable pumps to initiate make-up flow into the SFPs
through a seismic standpipe and standard fire hoses routed to the
SFPs or to a spray nozzle. These strategies will continue to be
required as a license condition. Considering the very low
probability of beyond-design-basis accidents affecting the SFP,
these diverse strategies provide defense-in-depth and time to
provide additional make-up or spray water to the SFP before the
onset of any postulated off-site radiological release.
It is not necessary to postulate both a loss of spent fuel pool
cooling in conjunction with a loss of spent fuel pool make-up, and
such an event is not postulated in UFSAR Section 15.7.3.8 related to
SFP boiling and is not credible given the number of diverse sources
of make-up and the time available to supply make-up.
As currently discussed in UFSAR 9.1.2.3, spent fuel pool boiling
also will not adversely affect the integrity of the SFPs. The
reinforced concrete temperature differences and gradients were
determined based on an inside face temperature of 230[emsp14][deg]F
(water temperature of 212[emsp14][deg]F and gamma heating of
18[emsp14][deg]F). That analysis indicates that the SFP walls have
sufficient structural capability to accommodate this thermal
loading.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
[[Page 69717]]
The proposed changes do not alter any design basis or safety
limits for the plant. The applicable limits are spent fuel clad
temperature and spent fuel pool level. The spent fuel cladding
temperature is assured by maintaining water level to support natural
circulation cooling within the spent fuel racks. Forced cooling
keeps evaporative losses and Fuel Handling Building environs within
nominal limits. Thus, the SSCs that support the design and safety
limits are limited to those that maintain inventory (Spent Fuel Pool
and related structural components (pool liner, structure, and racks)
and sufficient equipment to replace evaporative or other losses.
Complete loss of make-up is not credible given the existence of
numerous sources of make-up and the time available to provide make-
up. No changes to the pool and its structures are proposed and make-
up capability remains assured.
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: Walker A. Matthews, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, CA
91770.
NRC Branch Chief: Bruce Watson.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-
366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, GA
Date of amendment request: August 4, 2015. A publicly-available
version is in ADAMS under Accession No. ML15216A602.
Description of amendment request: The licensee describes the
application as follows: ``This amendment corrects an obvious
typographical error in the Unit 1 FOL [Facility Operating License], and
on page 5.0.17 of the Unit 2 TS [Technical Specification]. The Degraded
Voltage Protection license condition in Part 2.C of the Unit 1 FOL
(DPR-57) is currently listed as condition number 10, whereas it should
be listed as condition number 11. In addition, this paragraph should be
further indented to the right, to clarify that it's a third level
paragraph (i.e. level 2.C.11). In addition to the FOL change, this
amendment corrects an incorrect Unit number in Hatch Unit 2 TS page
5.0.17. This page was inadvertently sent and issued stating Unit 1 on
the bottom left, whereas it should clearly state Unit 2. Lastly, this
amendment adds the term STAGGERED TEST BASIS to the Definitions section
of the Unit 1 and Unit 2 TS. This term was removed from the TS and
moved to the Surveillance Frequency Control Program (SFCP) when the NRC
issued the TSTF-425 license amendment in [January 3,] 2012 to relocate
specific surveillance frequency requirements to a licensee controlled
program. This term, however, was reintroduced into Section 5 of the TS
as a defined term when Hatch adopted the Control Room Envelope
Habitability Program (TSTF-448) [in an amendment issued on August 29,
2014]. Since it's currently used as a defined term in Section 5 of the
TS, it needs to be included in the Definitions section of the TS.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment contains no technical changes; all
proposed changes are administrative. These changes are consistent
with the intent of what has already been approved by the Nuclear
Regulatory Commission (NRC).
There are no accidents affected by this change, and therefore no
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment contains no technical changes; all
proposed changes are administrative. These changes are consistent
with the intent of what has already been approved by the Nuclear
Regulatory Commission (NRC).
There are no accidents affected by this change, and therefore no
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment contains no technical changes; all
proposed changes are administrative. These changes are consistent
with the intent of what has already been approved by the Nuclear
Regulatory Commission (NRC).
There are no accidents affected by this change, and therefore no
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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Inverness Center
Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
III. 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 can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
[[Page 69718]]
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 17, 2014, as supplemented by
letter dated August 13, 2015.
Brief description of amendments: The amendments revised the Cyber
Security Plan (CSP) Milestone 8 full implementation date as set forth
in the CSP Implementation Schedule for the following plants: Kewaunee
Power Station; Millstone Power Station, Unit Nos. 2 and 3; North Anna
Power Station, Unit Nos. 1 and 2; and Surry Power Station, Unit Nos. 1
and 2.
Date of issuance: October 7, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 216, 323, 269, 276, 258, 286, and 286. A publicly-
available version is in ADAMS under Accession No. ML15245A482.
Documents related to these amendment are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-43, DPR-65, DPR-49,
NPF-4, NPF-7, DPR-32, and DPR-37: Amendments revised the Facility
Operating Licenses.
Date of initial notice in Federal Register: May 5, 2015 (80 FR
25718). The supplement letter dated August 13, 2015, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 7, 2015.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: September 10, 2015, as supplemented by
letters dated September 30 and October 20, 2015.
Brief description of amendment: The amendment approved a one-time
extension of the Technical Specification (TS) completion time
associated with the Division 2 Shutdown Service Water Subsystem from 72
hours to 7 days in support of maintenance activities.
Date of issuance: October 22, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No: 207. A publicly-available version is in ADAMS under
Accession No. ML15280A258; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-62: The amendment revised the
TSs and License.
Date of initial notice in Federal Register: September 18, 2015 (80
FR 56498).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 22, 2015.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373, LaSalle County
Station, Unit 1 and Unit 2, LaSalle County, Illinois
Date of amendment request: January 12, 2015.
Brief description of amendments: The amendments deleted the
limiting condition for operation (LCO) Note for Technical Specification
(TS) Section 3.5.1, ``ECCS [emergency core cooling system]--
Operating.'' The current Note allowed the licensee to consider the low
pressure coolant injection subsystem associated with the residual heat
removal system to be OPERABLE under specified conditions.
Date of issuance: October 14, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 217 and 203. A publicly-available version is in
ADAMS under Accession No. ML15244B410; documents related to this
amendment are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. NPF-11 and NPF-18: Amendments
revised the Facility Operating License and TSs.
Date of initial notice in Federal Register: March 31, 2015 (80 FR
17091).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 14, 2015.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: December 19, 2014, as supplemented by
letter dated June 26, 2015.
Brief description of amendment: This amendment revised the
technical specifications (TSs) to adopt performance-based Type C
testing for the reactor containment, which would allow for extended
test intervals for Type C valves, and corrects an editorial issue in
the TSs.
Date of issuance: October 9, 2015.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment No.: 288. A publicly-available version is in ADAMS under
Accession No. ML15239B293; documents related to this amendment are
listed in the Safely Evaluation enclosed with the amendment.
Facility Operating License No. NPF-3: Amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: March 31, 2015 (80 FR
17090), and July 7, 2015 (80 FR 38759). The supplemental letter dated
June 26, 2015, provided additional information that clarified the
application, did not expand the scope of the application as previously
noticed, and did not change the staff's proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 9, 2015.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: December 30, 2014.
Brief description of amendment: This amendment revises the
technical specification (TS) surveillance requirement for the frequency
to verify that each containment spray system nozzle is unobstructed
from every 10 years to an event-based frequency.
Date of issuance: October 20, 2015.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment No.: 289. A publicly-available version is in ADAMS under
Accession No. ML15251A046; documents related to this amendment
[[Page 69719]]
are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-3: Amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: March 31, 2015 (80 FR
17090).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 20, 2015.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: June 30, 2014, as supplemented March 27,
2015.
Brief description of amendment: The amendment revised the Humboldt
Bay Power Plant, Unit 3 License to approve the revised Emergency Plan.
Date of issuance: September 23, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 46. A publicly-available version is in ADAMS under
Accession No. ML15148A361; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-7: Amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: August 19, 2014 (79 FR
49109).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 23, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: July 22, 2015.
Brief description of amendment: The amendment revised Technical
Specification Section 6.0, ``Administrative Controls,'' by changing the
``Shift Supervisor'' title to ``Shift Manager.''
Date of issuance: October 15, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 202. A publicly-available version is in ADAMS under
Accession No. ML15208A029; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-12: Amendment revised
the Renewed Facility Operating License.
Date of initial notice in Federal Register: August 14, 2015 (80 FR
48924), as corrected by Federal Register notice dated August 20, 2015
(80 FR 50663).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 15, 2015.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: June 17, 2015, as supplemented by
letters dated July 14, August 3, August 28, September 3, and September
21, 2015.
Brief description of amendment: The amendment adopted new Technical
Specification (TS) 3.7.16, ``Component Cooling System (CCS)--
Shutdown,'' and TS 3.7.17, ``Essential Raw Cooling Water (ERCW)
System--Shutdown,'' and revised TS 3.3.2, ``Engineered Safety Feature
Actuation System (ESFAS) Instrumentation,'' and TS 3.4.6, ``RCS Loops-
MODE 4,'' to support dual-unit operation of WBN Units 1 and 2.
Date of issuance: October 20, 2015.
Effective date: As of the date of issuance and shall be implemented
after the issuance of the Facility Operating License for Unit 2.
Amendment No.: 104. A publicly-available version is in ADAMS under
Accession No. ML15275A042; documents related to this amendment are
listed in the Safety Evaluation (SE) enclosed with the amendment.
Facility Operating License No. NPF-90: Amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: July 17, 2015 (80 FR
42552). The supplemental letters dated July 14, August 3, August 28,
September 3, and September 21, 2015, provided additional information
that clarified the application. These supplements did not change the
staff's proposed no significant hazards consideration. The supplemental
letter dated September 3, 2015, provided additional information that
expanded the scope of the application as originally noticed. A notice
published in the Federal Register on September 15, 2015 (80 FR 55383),
supersedes the original notice in its entirety to update the expanded
scope of the amendment description and include the staff's proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in an SE dated October 20, 2015.
No significant hazards consideration determination comments
received: No.
Dated at Rockville, Maryland, this 2nd day of November, 2015.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-28347 Filed 11-9-15; 8:45 am]
BILLING CODE 7590-01-P