[Federal Register Volume 82, Number 1 (Tuesday, January 3, 2017)]
[Notices]
[Pages 154-164]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-31813]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0273]
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
[[Page 155]]
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 December 6 to December 19, 2016. The last
biweekly notice was published on December 20, 2016.
DATES: Comments must be filed by February 2, 2017. A request for a
hearing must be filed by March 6, 2017.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0273. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
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: Janet Burkhardt, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1384, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0273, facility name, unit
number(s), plant docket number, application date, and subject 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-2016-0273.
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 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-2016-0273, facility name, unit
number(s), plant docket number, 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 will post all comment submissions at http://www.regulations.gov as well as enter 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 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, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if 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. If the Commission
takes 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. If the Commission makes 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
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. 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/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
[[Page 156]]
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
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 petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) 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.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the 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) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
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 establish 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 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 the 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 March
6, 2017. 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 set forth
in this section, except that under 10 CFR 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.
Alternatively, a State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may participate as a non-party under 10
CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
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
his or her 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. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), 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 that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). 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. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. 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 (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (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. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit
[[Page 157]]
adjudicatory documents. Submissions should be in Portable Document
Format (PDF). Additional guidance on PDF submissions is available on
the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the
document is 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
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing 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 Electronic Filing Help Desk is available between 9 a.m.
and 7 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
stating why there is good cause for not filing electronically and
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, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents 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 personal phone numbers in their filings, unless an
NRC regulation or other law requires submission of such information.
For example, in some instances, individuals provide home addresses in
order to demonstrate proximity to a facility or site. With respect to
copyrighted works, except for limited excerpts that serve the purpose
of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
For further details with respect to 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.
Dominion Nuclear Connecticut, Inc. (DNC), Docket No. 50-336,
Millstone Power Station, Unit No. 2 (MPS2), New London County,
Connecticut.
Date of amendment request: December 14, 2016. A publicly-available
version is in ADAMS under Accession No. ML16354A424.
Description of amendment request: The amendment would revise
Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.2
regarding control element assembly (CEA) freedom of movement
surveillance, such that CEA 39 may be excluded from the last remaining
quarterly performance of the SR in MPS2 Cycle 24.
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 would exclude CEA 39 from SR 4.1.3.1.2
for the remainder of MPS2 Cycle 24 operation. The function of CEA 39
is to provide negative reactivity addition into the core upon
receipt of a signal from the Reactor Protection System (RPS). CEA 39
was demonstrated to be moveable and trippable during the last
performance of SR 4.1.3.1.2. Since the functionality of CEA 39 has
not been affected, the assumptions and conclusions of the Final
Safety Analysis Report (FSAR) Chapter 14, Safety Analysis, are not
affected by this license amendment request.
The misoperation of a CEA, which includes a CEA drop event, has
been evaluated in the MPS2 FSAR and found acceptable. The proposed
change would minimize the potential for inadvertent insertion of CEA
39 into the core by eliminating the requirement to place the CEA on
the UGC to perform freedom of movement testing. The proposed change
does not significantly increase the probability of a failure of a
CEA to insert into the core on a reactor trip or the probability of
an inadvertent CEA drop into the core at power.
No modifications are proposed to the RPS or associated Control
Element Drive Mechanism (CEDM) system logic.
Based on the above, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2
for the remainder of MPS2 Cycle 24 operation. CEA 39 was
demonstrated to be moveable and trippable during the last
performance of SR 4.1.3.1.2; therefore, the functionality of CEA 39
has not been affected. The proposed change will not introduce any
new design changes or systems that can prevent the CEA from
performing its specified safety function to insert on a reactor
trip. The current MPS2 FSAR safety analysis considers the drop of a
CEA into the core as an initiating event. This change does not alter
assumptions made in the FSAR Chapter 14 safety analysis.
Based on the above, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2
for the remainder of MPS2 Cycle 24 operation. SR 4.1.3.1.2 is
intended to verify freedom of movement of CEAs (i.e., trippable).
CEA 39 was demonstrated to be moveable and trippable during the last
performance of SR 4.1.3.1.2.
[[Page 158]]
The physical and electrical design of the CEAs, and past operating
experience, provides high confidence that CEAs remain trippable
whether or not exercised during each SR interval. Eliminating
further exercise of CEA 39 for the remainder of MPS2 Cycle 24
operation does not directly relate to the potential for CEA binding
to occur. The current MPS2 FSAR safety analysis is unaffected by
this license amendment request and there is no reduction in the
margin of safety.
There is no known failure mechanism (e.g., crud deposition) that
would preclude the CEA from inserting into core on a valid trip
signal or loss of power.
Based on the above, the proposed amendment does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Acting Branch Chief: Stephen S. Koenick.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
Date of amendment request: June 30, 2016. A publicly-available
version is in ADAMS under Accession No. ML16193A656.
Description of amendment request: The amendments would modify
Technical Specification 3.6.14, ``Divider Barrier Integrity,'' to
revise Conditions A and D to allow one steam generator (SG) enclosure
hatch or one pressurizer enclosure hatch to be open for up to 48 hours
to facilitate potential inspections and maintenance and to enhance
personnel and radiation safety.
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.
Implementation of this amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Removal of the SG enclosure hatch or the
pressurizer enclosure hatch will not cause an increase in the
probability of an accident that has been previously evaluated
because the hatches are not accident initiators.
The consequences of an accident, which have been previously
evaluated, will not be significantly increased by removal of the
pressurizer enclosure or SG enclosure hatch. As discussed in the
technical justification supporting this amendment request, the new
containment compression peak pressure will remain well below the
acceptance criteria. Additionally, the long term containment peak
pressure will not be adversely affected due to the delay time in
melting of the ice.
The removal of the pressurizer enclosure hatch itself has been
previously evaluated in Modes 1 through 4 in accordance with the
analytical method described in NUREG-0612 and the NRC's December 22,
1980, letter regarding the control of heavy loads at nuclear power
plants [(ADAMS Accession Nos. ML070250180 and ML071080219,
respectively)]. Because the SG enclosure hatch weighs less than 300
pounds, it would not be considered a heavy load as defined by NUREG-
0612. As such, it is not subject to heavy lift considerations.
Regardless, there is no safety-related equipment directly under
these hatch covers, so in the unlikely event that one fell, no
damage is expected to be caused. The changes proposed in this
[license amendment request (LAR)] have no adverse effect on the
procedures used for the handling of heavy loads at McGuire.
In summary, the proposed changes will not involve any 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.
Operation in accordance with the proposed amendment does not
create a new plant configuration and does not adversely affect how
the plant is operated, so implementation of this amendment would not
create the possibility of a new or different kind of accident from
any accident previously evaluated. No new accident causal mechanisms
are created as a result of the NRC approval of this license
amendment request. As discussed above, extending the time that the
pressurizer enclosure hatch or SG enclosure hatch is allowed to be
open or inoperable does not create any new or different accidents
from those previously evaluated. Removal of the pressurizer
enclosure hatch to perform inspections or maintenance has been
previously evaluated and determined to be acceptable. The analysis
contained in the technical justification for this license amendment
request provides results concluding that the containment compression
peak pressure and the long term containment peak pressure are
acceptable with either a pressurizer enclosure hatch or [an] SG
enclosure hatch open. This proposed amendment does not impact any
plant systems that are accident initiators; therefore, no new
accident types are being created.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The pressurizer enclosure hatch
and the SG enclosure hatch, as well as their performances, have a
direct impact on the containment boundary since peak containment
pressure due to an accident could be affected. However, the analysis
supporting this amendment request concludes that the containment
compression peak pressure and the long term containment peak
pressure continue to be acceptable with the increased time a single
hatch is open.
Therefore, the performance of the fission product barriers will
not be significantly impacted by implementation of this amendment,
and no safety margins will be significantly impacted.
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: Kate Nolan, Deputy General Counsel, Duke
Energy Corporation, 526 South Church Street--DEC45A, Charlotte, NC
28202.
NRC Branch Chief: Michael T. Markley.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North
Carolina.
Date of amendment request: September 6, 2016, as supplemented by
letter dated November 9, 2016. Publicly-available versions are in ADAMS
under Accession Nos. ML16257A410 and ML16330A504, respectively.
Description of amendment request: The amendment would revise the
Technical Specifications to support an expansion of the core power-flow
operating range (i.e., Maximum Extended Load Line Limit Analysis Plus
(MELLLA+)).
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 operation in the MELLLA+ operating domain does not
significantly increase the probability or consequences of an
accident previously evaluated. The probability (i.e., frequency of
occurrence) of
[[Page 159]]
Design Basis Accidents (DBAs) occurring is not affected by the
MELLLA+ operating domain because BSEP continues to comply with the
regulatory and design basis criteria established for plant
equipment. Furthermore, a probabilistic risk assessment demonstrates
that the calculated core damage frequencies do not significantly
change due to the MELLLA+.
There is no change in consequences of postulated accidents when
operating in the MELLLA+ operating domain compared to the operating
domain previously evaluated. The results of accident evaluations
remain within the NRC approved acceptance limits.
The spectrum of postulated transients has been investigated and
is shown to meet the plant's currently licensed regulatory criteria.
Continued compliance with the Safety Limit Minimum Critical Power
Ratio (SLMCPR) will be confirmed on a cycle-specific basis
consistent with the criteria accepted by the NRC.
Challenges to the reactor coolant pressure boundary were
evaluated for the MELLLA+ operating domain conditions (i.e.,
pressure, temperature, flow, and radiation) and were found to meet
their respective acceptance criteria for allowable stresses and
overpressure margin.
Challenges to the containment were evaluated and the containment
and its associated cooling systems continue to meet the current
licensing basis. The calculated post-Loss of Coolant Accident (LOCA)
suppression pool temperature remains acceptable.
The proposed changes to the sodium pentaborate (SPB) enrichment
and volume requirements maintain the capability of the Standby
Liquid Control (SLC) system to perform this reactivity control
function and ensure continued compliance with the requirements of 10
CFR 50.62. The SLC system is not considered to be an initiator of
any event. The use of the proposed SPB solution with a higher boron-
10 (B-10) isotope enrichment does not alter the design, function, or
operation of the SLC system or increase the likelihood of
malfunction that could increase the consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed operation in the MELLLA+ operating domain does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Equipment that could be affected by the MELLLA+ operating domain
has been evaluated. No new operating mode, safety-related equipment
lineup, accident scenario, or equipment failure mode was identified.
The full spectrum of accident considerations has been evaluated and
no new or different kind of accident was identified. The MELLLA+
operating domain uses developed technology, and applies it within
the capabilities of existing plant safety-related equipment in
accordance with the regulatory criteria, including NRC-approved
codes, standards and methods. The use of the proposed SPB solution
with a higher B-10 isotope enrichment does not alter the design,
function, or operation of the SLC system or create the possibility
of a new or different kind of accident. The proposed changes have
been assessed and determined not to introduce a different accident
than that previously evaluated. No new accident or event precursor
has been identified.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed operation in the MELLLA+ domain does not involve a
significant reduction in the margin of safety.
The MELLLA+ operating domain can only affect design and
operational margins. Challenges to the fuel, reactor coolant
pressure boundary, and containment were evaluated for the MELLLA+
operating domain conditions. Fuel integrity is maintained by meeting
existing design and regulatory limits. The calculated loads on
affected structures, systems, and components, including the reactor
coolant pressure boundary, will remain within their design
allowables for design basis event categories. No NRC acceptance
criterion is exceeded. The BSEP configuration and responses to
transients and postulated accidents do not result in exceeding the
presently approved NRC acceptance limits, thereby preserving safety
margins.
The proposed changes to the SPB enrichment and volume
requirements ensure SLC system shutdown margins and post-accident pH
control margins are maintained while maintaining compliance with the
requirements of 10 CFR 50.62.
Therefore, the proposed amendments do not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Jeanne D. Johnston.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos.
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3
(PBAPS), York and Lancaster Counties, Pennsylvania.
Date of amendment request: November 4, 2016, as supplemented by
letter dated December 7, 2016. Publicly-available versions are in ADAMS
under Accession Nos. ML16309A298 and ML16342C455, respectively.
Description of amendment request: The amendments would revise the
Allowable Value (AV) for the Turbine Condenser--Low Vacuum scram
function specified in Technical Specification Table 3.3.1.1-1,
``Reactor Protection System Instrumentation.'' The licensee stated that
the purpose of the proposed change is to minimize the potential for
inadvertent scrams due to low condenser vacuum.
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, with NRC edits shown in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change implements a revised AV [allowable value]
for the Turbine Condenser--Low Vacuum scram instrument function at
PBAPS.
The proposed change to the PBAPS Turbine Condenser- Low Vacuum
scram AV does not require modifying any system interface or affect
the probability of any event initiators at the facility. Overall RPS
[Reactor Protection System] performance will remain within the
bounds of the previously performed accident analyses, since [the
Turbine-Condenser--Low Vacuum scram is not specifically credited in
any accident analysis.]
There will be no degradation in the performance of, or an
increase in the number of challenges imposed on safety-related
equipment that are assumed to function during an accident situation.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in
the UFSAR [Updated Final Safety Analysis Report]. The proposed
change is consistent with 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 to the PBAPS Turbine Condenser--Low Vacuum
scram AV does not affect the design, functional performance, or
operation of the facility. Similarly, the proposed change does not
affect the design or operation of any SSCs [structures, systems, or
components] involved in the mitigation of any accidents, nor does it
affect the design or operation of any component in the facility such
that new equipment failure modes are created.
[[Page 160]]
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this change.
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.
The proposed change does not affect the acceptance criteria for
any analyzed event, nor is there a change to any Safety Analysis
Limit. There will be no effect on the manner in which safety limits,
limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions.
The purpose of the condenser low vacuum turbine trip is to
protect the main condenser against overpressure on loss of condenser
vacuum. A condenser low vacuum condition provides a signal to trip
the main turbine by providing automatic closure to the turbine stop
valves. To anticipate the transient and scram which results from the
closure of the turbine stop valves, a condenser low vacuum condition
initiates a reactor scram. The condenser low vacuum scram trip
setting is selected to initiate a reactor scram prior to initiation
of closure of the [t]urbine [s]top [v]alves.
The proposed LAR [license amendment request] does not change the
sequential relationship of the condenser low vacuum scram and
turbine trip. The Automatic Scram signal (Actual Trip Setpoint
greater than or equal to 21.95 inches [mercury (Hg)] vacuum) will
still occur prior to the Turbine Trip signal (Actual Trip Setpoint
20.0 inches Hg vacuum). This aligns with UFSAR Section 7.2 in that
the condenser low vacuum scram is an anticipatory trip prior to the
scram that would result from the closure of the main turbine stop
valves.
The condenser low vacuum scram is not specifically credited in
any accident analysis. The integrity of the condenser is not
compromised by the proposed change because the reactor will be shut
down using both diverse and redundant tripping to ensure fission
products are not released.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis. Based on this
review, and the NRC edits shown in square brackets, 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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL
60555.
NRC Acting Branch Chief: Stephen S. Koenick.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348
and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama.
Date of amendment request: November 22, 2016. A publicly-available
version is in ADAMS under Accession No. ML16336A024.
Description of amendment request: SNC requested to revise the
licensing basis that support a selected scope application of an
Alternative Source Term (AST) methodology and incorporate Technical
Specification Task Force (TSTF) Traveler, TSTF-448-A, Revision 3,
``Control Room Habitability,'' and TSTF-312-A, ``Administrative Control
of Containment Penetrations.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The licensee's analysis is
presented below, with NRC staff edits in square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
There are no physical changes to the plant being introduced by
the proposed changes to the accident source term. Implementation of
Alternative Source Term (AST) and the new atmospheric dispersion
factors have no impact on the probability for initiation of any
Design Basis Accidents (DBAs). Once the occurrence of an accident
has been postulated, the new accident source term and atmospheric
dispersion factors are an input to analyses that evaluate the
radiological consequences. The proposed changes do not involve a
revision to the design or manner in which the facility is operated
that could increase the probability of an accident previously
evaluated in Chapter 15 of the Final Safety Analysis Report (FSAR).
Based on the AST analyses, there are no proposed changes to
performance requirements and no proposed revision to the parameters
or conditions that could contribute to the consequences of an
accident previously discussed in Chapter 15 of the FSAR. Plant-
specific radiological analyses have been performed using the AST
methodology and new atmospheric dispersion factors (X/Qs) have been
established. Based on the results of these analyses, it has been
demonstrated that the Control Room and off-site dose consequences of
the limiting events considered in the analyses meet the regulatory
guidance provided for use with the AST, and the doses are within the
limits established by 10 CFR 50.67.
Regarding TSTF-312-A, the proposed change would allow
containment penetrations to be unisolated under administrative
controls during core alterations or movement of irradiated fuel
assemblies within containment. The status of containment penetration
flow paths (i.e., open or closed) is not an initiator for any design
basis accident or event, and therefore the proposed change does not
increase the probability of any accident previously evaluated. The
proposed change does not affect the design of the primary
containment, or alter plant operating practices such that the
probability of an accident previously evaluated would be
significantly increased. The proposed change does not significantly
change how the plant would mitigate an accident previously
evaluated, and is bounded by the fuel handling accident (FHA)
analysis.
Therefore, it is concluded that the proposed amendment does not
involve a significant increase in the probability or the
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.
No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any failure mode not
bounded by previously evaluated accidents. Implementation of AST and
the associated proposed Technical Specification changes and new X/Qs
have no impact to the initiation of any DBAs. These changes do not
affect the design function or modes of operation of structures,
systems and components in the facility prior to a postulated
accident. Since structures, systems and components are operated no
differently after the AST implementation, no new failure modes are
created by this proposed change. The AST change itself does not have
the capability to initiate accidents.
Regarding TSTF-312-A, allowing penetration flow paths to be open
is not an initiator for any accident. The proposed change to allow
open penetration flow paths will not affect plant safety functions
or plant operating practices such that a new or different accident
could be created. There are no design changes associated with the
proposed changes, and the change does not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed). The change does not alter assumptions made in
the safety analysis, and is consistent with the safety analysis
assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Consequently, the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The AST analyses have been performed using approved
methodologies to ensure that analyzed events are bounding and safety
[[Page 161]]
margin has not been reduced. The dose consequences of these limiting
events are within the acceptance criteria presented in 10 CFR 50.67.
Thus, by meeting the applicable regulatory limits for AST, there is
no significant reduction in a margin of safety.
Regarding TSTF-312-A, TS 3.9.3 provides measures to ensure that
the dose consequences of a postulated FHA inside containment are
minimized. The proposed change to LCO 3.9.3 will allow penetration
flow path(s) to be open during refueling operations under
administrative control. These administrative controls will provide
assurance that prompt closure of open penetrations flow paths can
and will be achieved in the event of an FHA inside containment, and
will minimize dose consequences. The proposed change does not affect
the safety analysis acceptance criteria for ay [an] analyzed event,
nor is there a change to any safety analysis limit. The proposed
change does not alter the manner in which safety limits, limiting
safety system settings or limiting conditions for operation are
deterred, not [nor] is here [there] any adverse effect on those
plant systems necessary to assure the accomplishment of protective
functions. The proposed change will not result in plant operation in
a configuration outside the design basis.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, the
proposed amendment does not involve a significant reduction in
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: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama.
Date of amendment request: November 15, 2016. A publicly-available
version is in ADAMS under Accession No. ML16320A540.
Description of amendment request: The proposed changes would revise
Technical Specification 5.5.17, ``Containment Leakage Rate Testing
Program.'' The revision would increase the existing Type A integrated
leakage rate test program test interval from 10 years to 15 years;
adopt an extension of the containment isolation valve leakage testing
(Type C) frequency from 60 months to 75 months; adopt the use of
American National Standards Institute/American Nuclear Society (ANSI/
ANS) 56.8-2002, ``Containment System Leakage Testing Requirements'';
and adopt a grace interval of 9 months for Type A, Type B, and Type C
leakage tests, in accordance with Nuclear Energy Institute (NEI) 94-01,
Revision 3-A.
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 activity involves the revision of Joseph M. Farley
Nuclear Plant (FNP), Units 1 and 2, Technical Specification (TS)
5.5.17, ``Primary Containment Leakage Rate Testing Program,'' to
allow the extension of the Type A integrated leakage rate test
(ILRT) containment test interval to 15 years, and the extension of
the Type C local leakage rate test (LLRT) interval to 75 months. The
current Type A test interval of 120 months (10 years) would be
extended on a permanent basis to no longer than 15 years from the
last Type A test. The current Type C test interval of 60 months for
selected components would be extended on a performance basis to no
longer than 75 months. Extensions of up to nine months (total
maximum interval of 84 months for Type C tests) are permissible only
for non-routine emergent conditions.
The proposed extensions do not involve either a physical change
to the plant or a change in the manner in which the plant is
operated or controlled. The containment is designed to provide an
essentially leak tight barrier against the uncontrolled release of
radioactivity to the environment for postulated accidents. As such,
the containment and the testing requirements invoked to periodically
demonstrate the integrity of the containment exist to ensure the
plant's ability to mitigate the consequences of an accident, and do
not involve the prevention or identification of any precursors of an
accident.
The change in Type A test frequency to once-per-fifteen years,
measured as an increase to the total integrated plant risk for those
accident sequences influenced by Type A testing, based on the
internal events probabilistic risk analysis (PRA) is 1.08E-02
person-rem/year for Unit 1 and 9.89 E-03 person-rem/year for Unit 2.
Electric Power Research Institute (EPRI) Report No. 1009325,
Revision 2-A states that a very small population is defined as an
increase of <= 1.0 person-rem per year or <= 1% of the total
population dose, whichever is less restrictive for the risk impact
assessment of the extended ILRT intervals. This is consistent with
the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for
Nuclear Energy Institute (NEI) 94-01 and EPRI Report No. 1009325.
Moreover, the risk impact when compared to other severe accident
risks is negligible.
Therefore, this proposed extension does not involve a
significant increase in the probability of an accident previously
evaluated.
In addition, as documented in NUREG-1493, ``Performance-Based
Containment Leak-Test Program,'' dated January 1995, Types B and C
tests have identified a very large percentage of containment leakage
paths, and the percentage of containment leakage paths that are
detected only by Type A testing is very small. The FNP Type A test
history supports this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and (2) time based. Activity-based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. The LLRT requirements and administrative controls such
as configuration management and procedural requirements for system
restoration ensure that containment integrity is not degraded by
plant modifications or maintenance activities. The design and
construction requirements of the containment combined with the
containment inspections performed in accordance with American
Society of Mechanical Engineers (ASME) Section XI, and TS
requirements serve to provide a high degree of assurance that the
containment would not degrade in a manner that is detectable only by
a Type A test. Based on the above, the proposed test interval
extensions do not significantly increase the consequences of an
accident previously evaluated.
The proposed amendment also deletes exceptions previously
granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to
allow one-time extensions of the ILRT test frequency for FNP. These
exceptions were for activities that would have already taken place
by the time this amendment is approved; therefore, their deletion is
solely an administrative action that has no effect on any component
and no impact on how the unit is operated.
Therefore, the proposed change does not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment to the TS 5.5.17, ``Containment Leakage
Rate Testing Program,'' involves the extension of the FNP Type A
containment test interval to 15 years and the extension of the Type
C test interval to 75 months. The containment and the testing
requirements to periodically demonstrate the integrity of the
containment exist to ensure the plant's ability to mitigate the
consequences of an accident do not involve any accident precursors
or initiators. The proposed change does not involve a physical
change to the plant (i.e., no new or different type of equipment
will be installed) or a change to the manner in which the plant is
operated or controlled.
The proposed amendment also deletes exceptions previously
granted under TS
[[Page 162]]
Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to allow one-time
extensions of the ILRT test frequency for FNP. These exceptions were
for activities that would have already taken place by the time this
amendment is approved; therefore, their deletion is solely an
administrative action that does not result in any change in how the
unit is operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment to TS 5.5.17 involves the extension of
the FNP Type A containment test interval to 15 years and the
extension of the Type C test interval to 75 months for selected
components. This amendment does not alter the manner in which safety
limits, limiting safety system set points, or limiting conditions
for operation are determined. The specific requirements and
conditions of the TS Containment Leak Rate Testing Program exist to
ensure that the degree of containment structural integrity and leak-
tightness that is considered in the plant safety analysis is
maintained. The overall containment leak rate limit specified by TS
is maintained.
The proposed change involves only the extension of the interval
between Type A containment leak rate tests and Type C tests for FNP.
The proposed surveillance interval extension is bounded by the 15-
year ILRT interval and the 75-month Type C test interval currently
authorized within NEI 94-01, Revision 3-A. Industry experience
supports the conclusions that Type B and C testing detects a large
percentage of containment leakage paths and that the percentage of
containment leakage paths that are detected only by Type A testing
is small. The containment inspections performed in accordance with
ASME Section XI and Technical Specifications serve to provide a high
degree of assurance that the containment would not degrade in a
manner that is detectable only by Type A testing. The combination of
these factors ensures that the margin of safety in the plant safety
analysis is maintained. The design, operation, testing methods and
acceptance criteria for Types A, B, and C containment leakage tests
specified in applicable codes and standards would continue to be
met, with the acceptance of this proposed change, since these are
not affected by changes to the Type A and Type C test intervals.
The proposed amendment also deletes an exception previously
granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to
allow one-time extensions of the ILRT test frequency for FNP. This
exception was for an activity that would have already taken place by
the time this amendment is approved; therefore, the deletion is
solely an administrative action and does not change how the unit is
operated and maintained. Therefore, there is no reduction in any
margin of safety.
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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of amendment request: October 11, 2016. A publicly-available
version is in ADAMS under Accession No. ML16286A553.
Description of amendment request: The amendment would revise the
Technical Specification (TS) requirements to reference and allow use of
the NRC-approved core reload methodologies described in Westinghouse
topical reports (TRs) WCAP-16045-P-A, ``Qualification of the Two-
Dimensional Transport Code PARAGON,'' WCAP-16045-P-A, Addendum 1-A,
``Qualification of the NEXUS Nuclear Data Methodology,'' and WCAP-
10965-P-A, Addendum 2-A, ``Qualification of the New Pin Power Recovery
Methodology,'' for the Callaway Plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed license amendment would revise TS 5.6.5.b to add
additional TR references for NRC-approved methodologies used in core
reload designs and the determination of core operating limits,
thereby specifically approving the use of these methodologies for
the Callaway Plant. The additional analytical methodologies are
improvements over the current methodologies in use at Callaway
Plant. The NRC staff reviewed and approved these methodologies and
concluded that these analytical methods are acceptable as a
replacement for the current analytical method.
This proposed license amendment does not involve any physical
changes to the Callaway Plant. Additionally, the core operating
limits determined using the proposed analytical methods will
continue to assure that the reactor operates safely. On that basis,
the proposed changes do not involve an increase in the probability
of an accident.
The proposed changes will not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended functions to mitigate the consequences of an initiating
event within the assumed acceptance limits and therefore, does not
increase the likelihood of any failure mechanisms or precursors to
transients or accidents postulated and analyzed in the Callaway
Plant FSAR [Final Safety Analysis Report]. Operation of the reactor
with core operating limits determined by use of the proposed
analytical methods does not increase the reactor power level, does
not increase the core fission product inventory, and does not change
any radiological release assumptions. The proposed changes will not
alter any accident analysis assumptions discussed in the FSAR, nor
do they involve any changes to the requirement for Callaway Plant to
operate within the power distribution limits and shutdown margins
required by the TS and within the assumptions of the safety analyses
described in the FSAR. Therefore the proposed methodology and TS
changes do not involve a significant increase in the consequences of
an accident.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides revised analytical methods for
determining core operating limits, and does not change any system
functions or requirements. Acceptance criteria required to be met
for analyzed core performance under normal, transient and accident
conditions are not being changed, as the core operating limits will
continue to be established in accordance with NRC-approved methods.
The change does not involve physical alteration of the plant, as no
new or different type of equipment will be installed. The change
does not alter assumptions made in the safety analyses, but ensures
that the core will operate within safe limits. Consequently, this
change does not create new failure modes or mechanisms, and no new
accident precursors are generated.
Therefore, it is concluded that this change does not create the
possibility the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes do not physically alter safety-
related systems, nor do they affect the way in which safety related
systems perform their functions. The setpoints at which protective
actions are initiated are not altered by the proposed changes. The
availability of equipment
[[Page 163]]
required to be available to actuate upon demand for mitigating an
analyzed event is unchanged by the proposed amendment. The proposed
analytical methodologies are an improvement that allows more
accurate modeling of core performance. The NRC has reviewed and
approved the additional methodologies for use in lieu of the current
methodology; thus, the margin of safety is not reduced due to this
change.
Therefore, it is concluded that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
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.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
Date of amendment request: September 26, 2013, as supplemented by
letters dated January 8, 2014; October 13, 2014; November 12, 2014;
December 12, 2014; January 26, 2015; February 27, 2015; March 13, 2015;
July 15, 2015; August 20, 2015; September 9, 2015; October 1, 2015;
January 14, 2016; April 26, 2016; September 29, 2016; and November 21,
2016.
Brief description of amendments: The amendments revised the
condition for the fire protection program (FPP) in Facility Operating
Licenses such that the FPP is now based on the requirements of 10 CFR
50.48(c), ``National Fire Protection Association Standard NFPA 805.''
Date of issuance: December 6, 2016.
Effective date: As of the date of issuance and shall be implemented
as stated in the revised License Condition 2.C(4).
Amendment Nos.: 291 and 270. A publicly-available version is in
ADAMS under Accession No. ML16077A135; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: February 19, 2014 (79
FR 9492). The supplemental letters dated January 8, 2014; October 13,
2014; November 12, 2014; December 12, 2014; January 26, 2015; February
27, 2015; March 13, 2015; July 15, 2015; August 20, 2015; September 9,
2015; October 1, 2015; January 14, 2016; April 26, 2016; September 29,
2016; and November 21, 2016, 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 proposal
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 6, 2016.
No significant hazards consideration comments received: No.
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida.
Date of application for amendment: July 28, 2015.
Brief description of amendment: The amendment incorporates into the
license the transfer of ownership, held by Seminole Electric
Cooperative, Inc. (SEC), in CR-3 to DEF. The transfer of ownership will
take place pursuant to the Settlement, Release and Acquisition
Agreement, dated April 30, 2015, wherein DEF will purchase the 1.6994
percent ownership share in CR-3 held by SEC, leaving DEF as the sole
remaining licensee for CR-3.
Date of issuance: November 30, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 251. A publicly-available version is in ADAMS under
Accession No. ML16293A200; documents related to this amendment are
listed in the Safety Evaluation enclosed with the letter dated August
10, 2016 (ADAMS Accession No. ML16173A022).
Facility Operating License No. DPR-72: This amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: September 29, 2015 (80
FR 58513), and January 4, 2016 (81 FR 98).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 10, 2016.
No significant hazards consideration comments received: No.
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida.
Date of application for amendment: August 27, 2015, as supplemented
by letters dated March 2, 2016, and July 14, 2016.
Brief description of amendment: The amendment approved the CR-3
Permanently Defueled Emergency Plan, and Permanently Defueled Emergency
Action Level Bases Manual, for the Independent Spent Fuel Storage
Installation.
Date of issuance: December 5, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 252. A publicly-available version is in ADAMS under
Accession No. ML16244A099; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
[[Page 164]]
Facility Operating License No. DPR-72: This amendment revised the
License.
Date of initial notice in Federal Register: November 10, 2015 (80
FR 69711).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 5, 2016.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades
Nuclear Plant, Van Buren County, Michigan.
Date of amendment request: March 3, 2016, as supplemented by letter
dated June 7, 2016.
Brief description of amendment: The amendment approves the
implementation of an alternate repair criteria (ARC) called C-star, for
the portion of the steam generator (SG) tubes within the cold-leg
tubesheet. In addition, the amendment clarifies the intent and improves
the wording of the technical specifications regarding the previously
incorporated ARC for the hot-leg side of the SG's tubesheet. This was
previously approved by letter dated May 31, 2007, and Amendment No.
225.
Date of issuance: December 19, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 261. A publicly-available version is in ADAMS under
Accession No. ML16300A030; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-20: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 2, 2016 (81 FR
50747).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 19, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc.; Georgia Power Company;
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia;
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia.
Date of amendment request: December 15, 2015, as supplemented by
letter dated April 11, 2016.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to incorporate risk-informed
requirements regarding selected Required Action end states.
Additionally, it modified TS Required Actions with a Note prohibiting
the use of Limiting Condition for Operation Applicability 3.0.4.a when
entering the preferred end state (Mode 3).
Date of issuance: December 19, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 281 (Unit No. 1); 225 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16257A724;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: February 16, 2016 (81
FR 7841). The supplemental letter dated April 11, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 2016.
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: February 23, 2016, as supplemented by
letter dated July 22, 2016.
Brief description of amendment: The amendment approved revisions to
the WBN Dual Unit Fire Protection Report and revised the associated
License Condition regarding the WBN fire protection program.
Date of issuance: December 12, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 108. A publicly-available version is in ADAMS under
Accession No. ML16307A013; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-90: Amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: May 10, 2016 (81 FR
28901). The supplemental letter dated July 22, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 12, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 21st day of December 2016.
For the Nuclear Regulatory Commission.
George A. Wilson,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2016-31813 Filed 12-30-16; 8:45 am]
BILLING CODE 7590-01-P