[Federal Register Volume 80, Number 197 (Tuesday, October 13, 2015)]
[Notices]
[Pages 61476-61492]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-25860]
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NUCLEAR REGULATORY COMMISSION
[NRC-2015-0236]
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 September 15 to September 28, 2015. The last
biweekly notice was published on September 29, 2015.
DATES: Comments must be filed by November 12, 2015. A request for a
hearing must be filed by December 14, 2015.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different
[[Page 61477]]
method for submitting comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0236. 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.
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-2015-0236 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-2015-0236.
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
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-2015-0236, 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, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's
[[Page 61478]]
right under the Act to be made a party to the proceeding; (3) the
nature and extent of the requestor's/petitioner's property, financial,
or other interest in the proceeding; and (4) the possible effect of any
decision or order which may be entered in the proceeding on the
requestor's/petitioner's interest. The petition must also identify the
specific contentions which the requestor/petitioner seeks to have
litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment 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.
B. Electronic Submissions (E-Filing)
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.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3), New London
County, Connecticut
Date of amendment request: June 30, 2015. A publicly-available
version is in ADAMS under Accession No. ML15183A022.
Description of amendment request: The amendments would revise the
MPS2 and MPS3 Final Safety Analysis Reports (FSARs) to: (1) Delete the
information pertaining to the severe line outage detection (SLOD)
special protection system; (2) update the description of the tower
structures associated with the four offsite transmission lines feeding
Millstone Power Station; and (3) describe how the current offsite power
source configuration and design satisfies the requirements of General
Design Criteria (GDC)-17, ``Electric Power Systems,'' and GDC-5,
``Sharing of Structures, Systems, and Components.'' The amendments also
request NRC approval of a new Technical Requirements Manual (TRM)
requirement, ``Offsite Line Power Sources,'' for MPS2 and MPS3. With
one offsite transmission line nonfunctional, the TRM requirement would
allow 72 hours to restore the nonfunctional line with a provision to
allow up to 14 days if specific TRM action requirements are met.
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 any accident previously
evaluated?
Response: No.
The post-modification configuration of the offsite 345 [kilovolt
(kV)] transmission system (four lines separately supported and SLOD
disabled) improves overall grid reliability and continues to meet
the requirements for two independent sources of offsite power (GDC-
17). Therefore, the post-modification configuration does not
significantly increase the probability or consequences of a loss of
offsite power event. Likewise, the associated proposed changes to
the MPS2 and MPS3 FSARs to document the revised 345 kV transmission
line tower design and disabling of SLOD, do not increase the
probability or consequences of an accident previously evaluated in
the FSARs.
The grid (offsite power) is by design, the preferred power
source for the affected units. The grid provides a reliable source
of power to MPS2 and MPS3 while the units are at power, in the event
of unit trips, and when the units are shut down for maintenance. New
TRM requirements are proposed that will maintain adequate defense in
depth to ensure grid reliability and stability are preserved.
A loss of offsite power event is an anticipated operational
occurrence. The proposed changes do not significantly increase the
probability of this event. Additionally, as described in Chapter 14
(MPS2) and Chapter 15 (MPS3), several events are assumed to occur
coincident with a loss of offsite power. Sufficient onsite power
sources are available to mitigate these events and ensure the
consequences of the existing analyses for these events remain
bounding.
The proposed new TRM requirements for offsite line power sources
will not change the plant design or design requirements. The design
criteria for the offsite power system remain unchanged. Therefore,
the safety analyses as documented in the MPS2 and MPS3 FSARs remain
unchanged. Temporary reductions in the number of offsite lines from
four to three, in accordance with the proposed TRM action
requirements, will not adversely affect offsite power system
availability in the event of a loss of either MPS2, MPS3, the
largest other unit on the grid, or the most critical transmission
line. Use of the proposed TRM requirements will not cause an
accident to occur and will not change how accident mitigation
equipment is operated. Allowing one offsite line to be nonfunctional
for up to 14 days does not increase the probability of any
previously evaluated accidents.
Therefore, the proposed changes to the offsite 345 kV
transmission system (four lines separately supported and SLOD
disabled) and proposed new TRM requirements does not significantly
increase the probability or consequences of an accident previously
evaluated.
[[Page 61479]]
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed amendments do not change the design function or
operation of the offsite power system and do not affect the offsite
power systems ability to perform its design function. The proposed
amendments do not conflict with the design criteria, codes, or
standards committed to in the licensing basis. The existing codes
and standards, as they apply to the onsite emergency power systems,
remain unchanged. The design criteria for the offsite power system
remain unchanged. Therefore, the safety analyses as documented in
the MPS2 and MPS3 FSARs remain unchanged.
No credible new failure mechanisms, malfunctions, or accident
initiators not considered in the design and licensing basis are
created by the proposed amendment. The offsite power system is
assumed to be available during several FSAR Chapter 14 (MPS2) and
Chapter 15 (MPS3) events. The new TRM requirements would allow 72
hours to restore a nonfunctional line, and up to 14 days to restore
a nonfunctional line if specific TRM action requirements are met.
Use of these TRM requirements does not impact offsite power
availability and does not create the possibility for a new or
different kind of accident from any previously evaluated. Temporary
reductions in the number of offsite lines from four to three, in
accordance with the proposed TRM requirements, will continue to
ensure offsite power system availability in the event of a loss of
either MPS2, MPS3, the largest other unit on the grid, or the most
critical transmission line.
The proposed amendments have no adverse effect on plant
operation or accident mitigation equipment. The response of the
plants and the operators following a design basis accident will not
be different. In addition, the proposed amendments do not create the
possibility of a new failure mode associated with any equipment or
personnel failures.
Therefore, the proposed amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The post-modification configuration of the offsite 345 kV
transmission system (four lines separately supported and SLOD
disabled) improves overall grid reliability and continues to meet
the requirements for two independent sources of offsite power (GDC-
17). Likewise, the addition of TRM requirements that limit the
unavailability of offsite lines provides acceptable assurance that
line outages will not result in a significant reduction to grid
stability and hence also to the margin of safety.
The offsite power systems are assumed to be available during
several FSAR Chapter 14 (MPS2) and Chapter 15 (MPS3) events. The
loss of the offsite power system is an anticipated operational
occurrence.
Additionally, as described in Chapter 14 (MPS2) and Chapter 15
(MPS3), several events are assumed to occur coincident with a loss
of offsite power. Sufficient onsite power sources are available to
mitigate these events and ensure the consequences of the existing
analyses for these events remain bounding.
The proposed amendments do not affect the assumptions in the
safety analyses or the ability to safely shutdown the reactors and
mitigate accident conditions. Station structures, systems, and
components will continue to be able to mitigate the design basis
accidents as assumed in the safety analyses and ensure proper
operation of accident mitigation equipment. In addition, the
proposed amendment will not affect equipment design or operation of
station structures, systems, and components and there are no changes
being made to the safety limits or safety system settings required
by technical specifications.
Therefore, the proposed amendments will not result in 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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Benjamin Beasley.
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: July 9, 2015. A publicly-available
version is in ADAMS under Accession No. ML15198A151.
Description of amendment request: The amendments would change the
reactor coolant pump (RCP) under-frequency trip setpoint Allowable
Value (AV) and add footnotes. The proposed license amendment request
affects Technical Specification (TS) 3.3.1, ``Reactor Trip System
Instrumentation,'' for McGuire Nuclear Station, Units 1 and 2.
Basis for proposed no significant hazards 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 TS changes involve lowering the existing RCP under-
voltage ALLOWABLE VALUE and adopting [Technical Specification Task
Force (TSTF)-493] provisions for as-found and as-left calibration
tolerances. The proposed TS changes serve to further ensure the
Reactor Trip RCP under-frequency and under-voltage trip
instrumentation will properly function as credited in the safety
analyses. The proposed changes do not alter any assumptions
previously made in the radiological consequences evaluations nor do
they affect mitigation of the radiological consequences of an
accident previously evaluated. The proposed TS changes do not affect
the probability of accident initiation.
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.
The proposed TS changes involve lowering the existing RCP under-
voltage ALLOWABLE VALUE and adopting TSTF-493 provisions for as-
found and as-left calibration tolerances. No new accident scenarios,
failure mechanisms, or single failures are introduced as a result of
any of the proposed changes.
The Reactor Trip System is not an accident initiator. No changes
to the overall manner in which the plant is operated are being
proposed.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant system
pressure boundary, and the containment barriers. The proposed TS
changes serve to ensure proper operation of the Reactor Trip RCP
under-frequency and under-voltage trip instrumentation and that the
instrumentation will properly function as credited in the safety
analyses. The proposed TS changes will not have any effect on the
margin of safety of fission product barriers. No accident mitigating
equipment will be adversely impacted as a result of the
modification.
Therefore, existing safety margins will be preserved. None of
the proposed changes will involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Robert J. Pascarelli.
[[Page 61480]]
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina; Docket
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2,
Mecklenburg County, North Carolina; and 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 15, 2015. A publicly-available
version is available at ADAMS Accession No. ML15196A093.
Description of amendment request: The proposed amendments would
revise the facilities Updated Final Safety Analysis Reports (UFSARs) to
provide gap release fractions for high-burnup fuel rods that exceed the
linear heat generation rate limit detailed in Table 3 of Regulatory
Guide (RG) 1.183, ``Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' July
2000 (ADAMS Accession No. ML003716792).
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 involves using gap release fractions for
high-burnup fuel rods (i.e., greater than 54 [gigawatt days per
metric ton unit (GWD/MTU)] that exceed the 6.3 [kiloWatt per foot
(kW/ft)] linear heat generation rate (LHGR) limit detailed in Table
3, Footnote 11 of RG 1.183. Increased gap release fractions were
determined and accounted for in the dose analysis for Catawba
Nuclear Station (CNS), Units 1 and 2; McGuire Nuclear Station (MNS),
Units 1 and 2; and Oconee Nuclear Station (ONS), Units 1, 2, and 3.
The dose consequence reported in each site's Updated Final Safety
Analysis Report (UFSAR) were reanalyzed for fuel handling-type
accidents only. Dose consequences were not reanalyzed for other non-
fuel-handling accidents since no fuel rod that is predicted to enter
departure from nuclear boiling (DNB) will be permitted to operate
beyond the limits of RG 1.183, Table 3, Footnote 11. The current NRC
requirements, as described in 10 CFR 50.67, specifies dose
acceptance criteria in terms of Total Effective Dose Equivalent
(TEDE). The revised dose consequence analysis for fuel handling-type
events at CNS, MNS, and ONS meet the applicable TEDE dose acceptance
criteria (specified also in RG 1.183). A slight increase in dose
consequences is exhibited. However, the increase is not significant
and the new TEDE results are below regulatory acceptance criteria.
The changes proposed do not affect the precursors for fuel
handling-type accidents analyzed in Chapter 15 of the CNS, MNS, or
ONS UFSARs. The probability remains unchanged since the accident
analyses performed and discussed in the basis for the UFSAR changes,
involve no change to a system, structure, or component that affects
initiating events for any UFSAR Chapter 15 accident evaluated.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
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 involves using gap release fractions for
high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed
the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG
1.183. Increased gap release fractions were determined and accounted
for in the dose analysis for CNS, MNS, and ONS. The dose
consequences reported in each site's UFSAR were reanalyzed for fuel
handling-type accidents only. Dose consequences were not reanalyzed
for other non-fuel-handling accidents since no fuel rod that is
predicted to enter departure from nucleate boiling (DNB) will be
permitted to operate beyond the limits of RG 1.183, Table 3,
Footnote 11.
The proposed change does not involve the addition or
modification of any plant equipment. The proposed change has the
potential to affect future core designs for CNS, MNS, and ONS.
However, the impact will not be beyond the standard function
capabilities of the equipment. The proposed change involves using
gap release fractions that would allow high-burnup fuel rods (i.e.,
greater than 54 GWD/MTU) to exceed the 6.3 kW/ft LHGR limit detailed
in Table 3, Footnote 11 of RG 1.183. Accounting for these new gap
release fractions in the dose analysis for CNS, MNS, and ONS does
not create the possibility of a new accident.
Therefore, the proposed change does no 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 change involves using gap release fractions for
high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed
the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG
1.183. Increased gap release fractions were determined and accounted
for in the dose analysis for CNS, MNS, and ONS. The dose
consequences reported in each site's UFSAR were reanalyzed for fuel
handling-type accidents only. Dose consequences were not reanalyzed
for other non-fuel-handling accidents since no fuel rod that is
predicted to enter departure from nucleate boiling (DNB) will be
permitted to operate beyond the limits of RG 1.183, Table 3,
Footnote 11.
The proposed change has the potential for an increased
postulated accident dose at CNS, MNS or ONS. However, the analysis
demonstrates that the resultant doses are within the appropriate
acceptance criteria. The margin of safety, as described by 10 CFR
50.67 and Regulatory Guide 1.183, has been maintained. Furthermore,
the assumptions and input used in the gap release and dose
consequences calculations are conservative. These conservative
assumptions ensure that the radiation doses calculated pursuant to
Regulatory Guide 1.183 and cited in this license amendment requires
are the upper bounds to radiological consequences of the fuel
handling-type accidents analyzed. The analysis shows that with
increased gap release fractions accounted for in the dose
consequences calculations there is margin between the offsite
radiation doses calculated and the dose limits of 10 CFR 50.67 and
acceptance criteria of Regulatory Guide 1.183. The proposed change
will not degrade the plant protective boundaries, will not cause a
release of fission products to the public and will not degrade the
performance of any structures, systems and components important to
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: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAF), Oswego County, New York
Date of amendment request: August 20, 2015. A publicly-available
version is in ADAMS under Accession No. ML15232A761.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.5.6, ``Primary Containment Leak Rate
Testing Program,'' to allow permanent extension of the Type A Primary
Containment Integrated Leak Rate Test (ILRT) interval to 15 years and
to allow extension of Type C Local Leak Rate Test (LLRT) testing
interval up to 75 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or
[[Page 61481]]
consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the TS involves the extension of the
JAF Type A containment test interval to 15 years and the extension
of the Type C test 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 extension does 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 dose risk for changing
the Type A test frequency from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated plant
risk for those accident sequences influenced by Type A testing, is
0.0087 person-[roentgen equivalent man (rem)]/year. [Electric Power
Research Institute (EPRI)] Report No. 1009325, Revision 2-A states
that a very small population dose 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. The results of the risk assessment for this
amendment meet these criteria. Moreover, the risk impact for the
ILRT extension 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.
As documented in NUREG-1493 [``Performance Based Containment
Leak-Test Program''], Type 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 JAF 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. Local leak rate test 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)
Boiler and Pressure Vessel Code] Section XI, the Maintenance Rule,
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 extensions
do not significantly increase the consequences of an accident
previously evaluated.
The proposed amendment also deletes exceptions previously
granted to allow one-time extensions of the ILRT test frequency for
JAF. 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 involves the extension of the
JAF 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 to allow one-time extensions of the ILRT test frequency for
JAF. 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 5.5.6 involves the extension of the
JAF 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 JAF.
The proposed surveillance interval extension is bounded by the 15-
year ILRT Interval and the 75-month Type C test interval currently
authorized within [Nuclear Energy Institute (NEI) 94-01, Revision 3-
A [``Industry Guideline for Implementing Performance-Based Option of
10 CFR Part 50, Appendix J,'' July 2012 (ADAMS Accession No.
ML12221A202)]. Industry experience supports the conclusion 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 Xl, TS and the
Maintenance Rule 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 Type 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 exceptions previously
granted to allow one time extensions of the ILRT test frequency for
JAF. 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 and does not change how
the unit is operated and maintained. Thus, 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: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Benjamin G. Beasley.
Entergy Operations, Inc.; System Energy Resources, Inc.; South
Mississippi Electric Power Association; and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of amendment request: June 29, 2015. A publicly-available
version is in ADAMS under Accession No. ML15180A376.
Description of amendment request: The amendment proposes a change
to
[[Page 61482]]
the GGNS Cyber Security Plan (CSP) Milestone 8 full implementation date
as set forth in the CSP Implementation Schedule.
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 CSP Implementation Schedule is
administrative in nature. This change does not alter accident
analysis assumptions, add any initiators, or affect the function of
plant systems or the manner in which systems are operated,
maintained, modified, tested, or inspected. The proposed change does
not require any plant modifications which affect the performance
capability of the structures, systems and components relied upon to
mitigate the consequences of postulated accidents and has no impact
on the probability or consequences of an accident previously
evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the CSP Implementation Schedule is
administrative in nature. This proposed change does not alter
accident analysis assumptions, add any initiators or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems, and components
relied upon to mitigate the consequences of postulated accidents and
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
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.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed change to
the CSP Implementation Schedule is administrative in nature. In
addition, the milestone date delay for full implementation of the
CSP has no substantive impact because other measures have been taken
which provide adequate protection during this period of time.
Because there is no change to established safety margins as a result
of this change, the proposed change does not involve a significant
reduction in a 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: Joseph A. Aluise, Associate General
Counsel--Legal, Nuclear and Environmental, Entergy Services, Inc., 639
Loyola Avenue, New Orleans, LA 70113.
NRC Branch Chief: Meena K. Khanna.
Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear
Power Plant (Ginna), Wayne County, New York
Date of amendment request: June 4, 2015. A publicly-available
version is in ADAMS under Accession No. ML15166A075.
Description of amendment request: The amendment would modify
Ginna's technical specifications (TS) by relocating specific
surveillance frequencies to a licensee-controlled program with the
implementation of Nuclear Energy Institute (NEI) 04-10, [Rev. 1,
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed
Method for Control of Surveillance Frequencies,'' April 2007 (ADAMS
Accession No. ML071360456)].
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 any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program [SFCP]. Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly increased.
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?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
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 change involve a significant reduction in a
margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components, specified in applicable
codes and standards (or alternatives approved for use by the NRC)
will continue to be met as described in the plant licensing basis
(including the final safety analysis report and bases to TS), since
these are not affected by changes to the surveillance frequencies.
Similarly, there is no impact to safety analysis acceptance criteria
as described in the plant licensing basis. To evaluate a change in
the relocated surveillance frequency, Exelon will perform a
probabilistic risk evaluation using the guidance contained in NRC
approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04-
10, Rev. 1, methodology provides reasonable acceptance guidelines
and methods for evaluating the risk increase of proposed changes to
surveillance frequencies consistent with Regulatory Guide 1.177
[``An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications''].
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: J. Bradley Fewell, Senior Vice President,
Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation
Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Benjamin G. Beasley.
[[Page 61483]]
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2 (SL-1 and 2), St. Lucie County,
Florida
Date of amendment request: March 10, 2015. A publicly-available
version is in ADAMS under Accession No. ML15084A141.
Description of amendment request: The amendments would remove
Technical Specification (TS) Limiting Condition for Operation (LCO) 3/
4.9.5, ``Communications,'' from the SL-1 and 2 TSs; remove LCO 3/4.9.6,
``Manipulator Crane Operability,'' from the SL-1 TSs; and remove LCO 3/
4.9.6, ``Manipulator Crane,'' from the SL-2 TSs. Each of these TS
requirements will be relocated to the Updated Final Safety Analysis
Report (UFSAR) for SL-1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes act to remove the current necessity of
establishing and maintaining communications between the control room
and the refueling station and the minimum load capacities and load
limit controls required for the manipulator crane limits and
relocate the requirements to the UFSAR, which will have no impact on
any safety related structures, systems or components. Once relocated
to the UFSAR, changes to establishing and maintaining communications
between the control room and the refueling station and the minimum
load capacities and load limit controls required for the manipulator
crane limits will be controlled in accordance with 10 CFR 50.59.
The probability of occurrence of a previously evaluated accident
is not increased because these changes do not introduce any new
potential accident initiating conditions. The consequences of
accidents previously evaluated in the UFSAR are not affected because
the ability of the components to perform their required functions is
not affected.
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 changes act to remove the current necessity of
establishing and maintaining communications between the control room
and the refueling station and the minimum load capacities and load
limit controls required for the manipulator crane limits and
relocate the requirements to the UFSAR, which will have no impact on
any safety related structures, systems or components. Once relocated
to the UFSAR, changes to establishing and maintaining communications
between the control room and the refueling station and the minimum
load capacities and load limit controls required for the manipulator
crane limits will be controlled in accordance with 10 CFR 50.59.
The proposed changes do not introduce new modes of plant
operation and do not involve physical modifications to the plant (no
new or different type of equipment will be installed). There are no
changes in the method by which any safety related plant structure,
system, or component (SSC) performs its specified safety function.
As such, the plant conditions for which the design basis accident
analyses were performed remain valid.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of the proposed changes. There will be no adverse effect or
challenges imposed on any SSC as a result of the proposed changes.
Therefore, the proposed changes do 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.
Margin of safety is related to confidence in the ability of the
fission product barriers to perform their accident mitigation
functions. The proposed changes act to remove the current necessity
of establishing and maintaining communications between the control
room and the refueling station and the minimum load capacities and
load limit controls required for the manipulator crane limits and
relocate the requirements to the UFSAR, which will have no impact on
any safety related structures, systems or components. Once relocated
to the UFSAR, changes to establishing and maintaining communications
between the control room and the refueling station and the minimum
load capacities and load limit controls required for the manipulator
crane limits will be controlled in accordance with 10 CFR 50.59. The
proposed changes do not physically alter any SSC. There will be no
effect on those SSCs necessary to assure the accomplishment of
protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio (DNBR) limits, loss of
cooling accident peak cladding temperature (LOCA PCT), or any other
margin of safety. The applicable radiological dose consequence
acceptance criteria will continue to be met.
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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Shana R. Helton.
Northern States Power Company--Minnesota Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: September 2, 2015. A publicly-available
version is in ADAMS under Accession No. ML15246A530.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core
Cooling System]--Operating,'' to correct the current non-conservative
value specified for minimum Alternate Nitrogen System pressure. The
proposed change would revise the TS surveillance requirement (SR)
3.5.1.3.b pressure limit for determining operability of the Alternate
Nitrogen System from greater than or equal to (>=) 410 pounds per
square inch gauge (psig) to a corrected value of >=1060 psig.
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 provided 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 TS SR for the purpose of
restoring a value to be consistent with the licensing basis. The
proposed TS change does not introduce new equipment or new equipment
operating modes, nor does the proposed change alter existing system
relationships. The proposed change does not affect plant
operation[.] Further, the proposed change does not increase the
likelihood of the malfunction of any SSC [structure, system or
component] or impact any analyzed accident. Consequently, the
probability of an accident previously evaluated is not affected and
there is no significant increase in the consequences of any accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the TS SR for the purpose of
restoring a value to be consistent with the licensing basis. The
[[Page 61484]]
change does not involve a physical alteration to the plant (i.e., no
new or different type of equipment will be installed) or a change in
the methods governing normal plant operations. The proposed change
does not alter assumptions made in the safety analysis for the
components supplied by the Alternate Nitrogen System. Further, the
proposed change does not introduce new accident initiators.
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 revises the TS SR for the purpose of
restoring a value to be consistent with the licensing basis. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis assumptions and
acceptance criteria are not affected by this change.
Therefore, the proposed change 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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David L. Pelton.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: July 15, 2015. A publicly-available
version is in ADAMS under Accession No. ML15196A576.
Description of amendment request: The proposed amendment would
revise or add technical specification (TS) surveillance requirements
(SRs) that require verification that the Emergency Core Cooling System
(ECCS), the Residual Heat Removal (RHR) System/Shutdown Cooling (SDC)
System, the Containment Spray (CS) System, and the Reactor Core
Isolation Cooling (RCIC) System are not rendered inoperable due to gas
accumulation and to provide allowances which permit performance of the
revised verification. The changes are being made to address the
concerns discussed in NRC Generic Letter 2008-01, ``Managing Gas
Accumulation in Emergency Core Cooling, Decay Heat Removal, and
Containment Spray Systems.'' The proposed changes are based on Revision
2 of NRC-approved Technical Specification Task Force (TSTF) Traveler
TSTF-523, ``Generic Letter 2008-01, Managing Gas Accumulation,'' dated
February 21, 2013 (ADAMS Accession No. ML13053A075). The NRC staff
issued a Notice of Availability for TSTF-523, Revision 2, for plant-
specific adoption using the consolidated line item improvement process,
in the Federal Register on January 15, 2014 (79 FR 2700).
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 provided 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 or adds Surveillance Requirements
(SRs) that require verification that the Emergency Core Cooling
Systems (ECCS), the Residual Heat Removal (RHR) System/Shutdown
Cooling (SDC) System, the Containment Spray (CS) System, and the
Reactor Core Isolation Cooling (RCIC) System are not rendered
inoperable due to accumulated gas and to provide allowances which
permit performance of the revised verification. Gas accumulation in
the subject systems is not an initiator of any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The proposed SRs ensure
that the subject systems continue to be capable to perform their
assumed safety function and are not rendered inoperable due to gas
accumulation. Thus, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, the RHR/SDC System, the CS System, and
the RCIC System are not rendered inoperable due to accumulated gas
and to provide allowances which permit performance of the revised
verification. The proposed change does not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the proposed change does not impose any new
or different requirements that could initiate an accident. The
proposed change does not alter assumptions made in the safety
analysis and is consistent with the safety analysis assumptions.
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 revises or adds SRs that require
verification that the ECCS, the RHR/SDC System, the CS System, and
the RCIC System are not rendered inoperable due to accumulated gas
and to provide allowances which permit performance of the revised
verification. The proposed change clarifies requirements for
management of gas accumulation in order to ensure the subject
systems are capable of performing their assumed safety functions.
The proposed SRs are more comprehensive than the current SRs and
will ensure that the assumptions of the safety analysis are
protected. The proposed change does not adversely affect any current
plant safety margins or the reliability of the equipment assumed in
the safety analysis. Therefore, there are no changes being made to
any safety analysis assumptions, safety limits or limiting safety
system settings that would adversely affect plant safety as a result
of the 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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David L. Pelton.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: June 29, 2015. A publicly-available
version is in ADAMS under Accession No. ML15187A259.
Description of amendment request: The proposed amendment would
revise or add technical specification (TS) surveillance requirements
(SRs) that require verification that the Emergency Core Cooling System
(ECCS), the Residual Heat Removal (RHR) System, and the Containment
Spray (CS) System are not rendered inoperable due to gas accumulation
and to provide allowances which permit performance of the revised
verification. The changes are being made to address the concerns
discussed in NRC Generic Letter 2008-01, ``Managing Gas Accumulation in
Emergency Core Cooling, Decay Heat Removal, and Containment Spray
Systems.'' The proposed changes are
[[Page 61485]]
based on Revision 2 of NRC-approved Technical Specification Task Force
(TSTF) Traveler TSTF-523, ``Generic Letter 2008-01, Managing Gas
Accumulation,'' dated February 21, 2013 (ADAMS Accession No.
ML13053A075). The NRC staff issued a Notice of Availability for TSTF-
523, Revision 2, for plant-specific adoption using the consolidated
line item improvement process, in the Federal Register on January 15,
2014 (79 FR 2700).
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 provided 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 revises or adds Surveillance Requirements
(SRs) that require verification that the Emergency Core Cooling
System (ECCS), the Residual Heat Removal (RHR) System, and the
Containment Spray (CS) System are not rendered inoperable due to
accumulated gas and to provide allowances which permit performance
of the revised verification. Gas accumulation in the subject systems
is not an initiator of any accident previously evaluated. As a
result, the probability of any accident previously evaluated is not
significantly increased. The proposed SRs ensure that the subject
systems continue to be capable to perform their assumed safety
function and are not rendered inoperable due to gas accumulation.
Thus, the consequences of any accident previously evaluated are not
significantly increased.
Therefore, the proposed licensing basis 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 [revises or] adds SRs that require
verification that the ECCS, the RHR System, and the CS System are
not rendered inoperable due to accumulated gas and to provide
allowances which permit performance of the revised verification. The
proposed change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the proposed change does not impose any new or different
requirements that could initiate an accident. The proposed change
does not alter assumptions made in the safety analysis and is
consistent with the safety analysis assumptions.
Therefore, the proposed licensing basis 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 [revises or] adds SRs that require
verification that the ECCS, the RHR System, and the CS System are
not rendered inoperable due to accumulated gas and to provide
allowances which permit performance of the revised verification. The
proposed change adds new requirements to manage gas accumulation in
order to ensure the subject systems are capable of performing their
assumed safety functions. The proposed SRs will ensure that the
assumptions of the safety analysis are protected. The proposed
change does not adversely affect any current plant safety margins or
the reliability of the equipment assumed in the safety analysis.
Therefore, there are no changes being made to any safety analysis
assumptions, safety limits[,] or limiting safety system settings
that would adversely affect plant safety as a result of the proposed
change.
Therefore, the proposed licensing basis 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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David L. Pelton.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 24, 2015. A publicly-available
version is in ADAMS under Accession No. ML15205A276.
Description of amendment request: The amendment would revise the
Technical Specification (TS) Surveillance Requirements (SRs), which
currently require operating ventilation systems with charcoal filters
for a 10-hour period at a monthly frequency. The SRs would be revised
to require operation of the systems for 15 continuous minutes at a
monthly frequency. The proposed amendment is consistent with NRC-
approved Technical Specifications Task Force (TSTF) Traveler TSTF-522,
Revision 0, ``Revise Ventilation System Surveillance Requirements to
Operate for 10 hours per Month,'' as published in the Federal Register
on September 20, 2012 (77 FR 58428), with variations due to plant-
specific nomenclature. The changes would revise TS 3.2, Table 3-5; SR
Items 10a.3.a, ``Control Room Air Filtration System (CRAFS)''; 10b.3.a,
``Spent Fuel Pool Storage Area Filtration System (SFPSAFS)''; and
10c.3.a, ``Safety Injection Pump Room Air Filtration System
(SIPRAFS),'' and TS 3.6(3)c, ``Containment Recirculating Air Cooling
and Filtering System,'' also known as the Containment Air Cooling and
Filtering System (CACFS).
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 replaces an existing SR to operate the CRAFS
for ten (10) continuous hours every month with heaters operating
with a requirement to operate the system for 15 continuous minutes
every month with heaters operating. The proposed change also
replaces existing SRs to operate the SFPSAFS, the SIPRAFS, and the
CACFS for ten (10) hours every month with a requirement to operate
these systems for 15 continuous minutes every month.
These systems are not accident initiators and therefore, these
changes do not involve a significant increase in the probability of
an accident. The proposed system and filter testing changes are
consistent with current regulatory guidance for these systems. The
proposed changes continue to ensure that these systems perform their
design function, which may include mitigating accidents. Thus, the
change does 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 replaces an existing SR to operate the CRAFS
for ten (10) continuous hours every month with heaters operating
with a requirement to operate the system for 15 continuous minutes
every month with heaters operating. The proposed change also
replaces existing SRs to operate the SFPSAFS, the SIPRAFS, and the
CACFS for ten (10) hours every month with a requirement to operate
these systems for 15 continuous minutes every month.
The change proposed for these ventilation systems does not
change any system operations or maintenance activities. Testing
requirements will be revised and will continue to demonstrate that
the Limiting Conditions for Operation are met and/or the system
components are capable of performing their intended safety
functions. The 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 of a new or
[[Page 61486]]
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 replaces an existing SR to operate the CRAFS
for ten (10) continuous hours every month with heaters operating
with a requirement to operate the system for 15 continuous minutes
every month with heaters operating. The proposed change also
replaces existing SRs to operate the SFPSAFS, the SIPRAFS, and the
CACFS for ten (10) hours every month with a requirement to operate
these systems for 15 continuous minutes every month.
The design basis for the CRAFS heaters is to heat the incoming
air, which reduces the relative humidity. The heater testing change
proposed for the CRAFS will continue to demonstrate that the heaters
are capable of heating the air and will perform their design
function. The SFPSAFS, and the SIPRAFS are tested for adsorption at
a relative humidity of [95 percent (%)] in accordance with RG
[Regulatory Guide] 1.52, Revision 3, and do not require heaters for
these systems to perform their specified safety function. The CACFS
does not need to be tested similarly because the CACFS charcoal
filters are not credited for the removal of radioiodines. The
proposed change is consistent with regulatory guidance.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: August 20, 2015. A publicly-available
version is in ADAMS under Accession No. ML15233A494.
Description of amendment request: The amendment would make
administrative changes to update personnel and committee titles in the
Technical Specifications (TSs), delete outdated or completed additional
actions contained in Appendix B of the license, and relocate the
definition of Process Control Program from the TSs to the Updated
Safety Analysis Report (USAR). The changes are proposed by the licensee
to use consistent terminology with Exelon Generation Company as part of
their Operating Services Agreement.
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 changes are administrative in nature, involving
changes to personnel and committee titles, deletion and or re-
location of requirements redundant to regulations, and deletion of
conditions controlling the first performance of testing that has
since been completed. The proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated because: (1) the proposed amendment
does not represent a change to the system design, (2) the proposed
amendment does not alter, degrade, or prevent action described or
assumed in any accident in the USAR from being performed, (3) the
proposed amendment does not alter any assumptions previously made in
evaluating radiological consequences, and [(4)] the proposed
amendment does not affect the integrity of any fission product
barrier. No other safety related equipment is affected by the
proposed change.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Hence, the proposed changes do not introduce any new
accident initiators, nor do these changes reduce or adversely affect
the capabilities of any plant structure or system in the performance
of their safety function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits or limiting safety system settings are determined. The safety
analysis acceptance criteria are not affected by these proposed
changes. Further, the proposed changes do not change the design
function of any equipment assumed to operate in the event of an
accident.
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: David A. Repka, Esq., Winston & Strawn, 1700
K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment request: June 17, 2015, as supplemented by letter
dated August 31, 2015. Publicly-available versions are in ADAMS under
Accession Nos. ML15176A539 and ML15243A363, respectively.
Description of amendment request: The amendments would revise the
licensing bases to adopt the alternative source term (AST) as allowed
by 10 CFR 50.67, ``Accident source term.'' The AST methodology, as
established in NRC Regulatory Guide (RG) 1.183, ``Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors,'' July 2000 (ADAMS Accession No. ML003716792),
is used to calculate the offsite and control room radiological
consequences of postulated accidents for DCPP, Unit Nos. 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This license amendment does not physically impact any system,
structure, or component (SSC) that is a potential initiator of an
accident. Therefore, implementation of AST, the AST assumptions and
inputs, the proposed [Technical Specification (TS)] changes, and new
[chi]/Q values have no impact on the probability for initiation of
any design basis accident. Once the occurrence of an accident has
been postulated, the new accident source term and [atmospheric
dispersion factors ([chi]/Q)] values are inputs to analyses that
evaluate the radiological consequences of the postulated events.
Reactor coolant specific activity, testing criteria of charcoal
filters, and the accident induced primary-to-secondary system
leakage performance criterion are not initiators for any accident
previously evaluated. The proposed change to require the 48-inch
containment purge valves to be sealed closed during operating MODES
1, 2, 3, and 4 is not an accident initiator for any
[[Page 61487]]
accident previously evaluated. The change in the classifications of
a portion of the 40-inch Containment Penetration Area Ventilation
line and a portion of the 2-inch gaseous radwaste system line is
also not an accident initiator for any accident previously
evaluated. Thus, the proposed TS changes and AST implementation will
not increase the probability of an accident.
The change to the decay time prior to fuel movement is not an
accident initiator. Decay time is used to determine the source term
for the dose consequence calculation following a potential [fuel
handling accident (FHA)] and has no effect on the probability of the
accident. Likewise, the change to the Control Room radiation
monitors setpoint cannot cause an accident and the operation of
containment spray during the recirculation phase is used for
mitigation of a [loss-of-coolant accident (LOCA)], and thus not an
accident initiator.
As a result, there are no proposed changes to the parameters or
conditions that could contribute to the initiation of an accident
previously evaluated in Chapter 15 of the Updated Final Safety
Analysis Report (UFSAR). As such, the AST cannot affect the
probability of an accident previously evaluated.
Regarding accident consequences, equipment and components
affected by the proposed changes are mitigative in nature and relied
upon once the accident has been postulated. The license amendment
implements a new calculation methodology for determining accident
consequences and does not adversely affect any plant component or
system that is credited to mitigate fuel damage. Subsequently, no
conditions have been created that could significantly increase the
consequences of any accidents previously evaluated.
Requiring that the 48-inch containment purge supply and exhaust
valves be sealed closed during operating MODES 1, 2, 3, and 4
eliminates a potential path for radiological release following
events that result in radioactive material releases to the
containment, thus reducing potential consequences of the event. The
steam generator tube inspection testing criterion for accident
induced leakage is being changed, resulting in lower leakage rates,
and thus less potential releases due to primary-to-secondary
leakage. The auxiliary building ventilation system allowable methyl
iodide penetration limit is being changed, which results in more
stringent testing requirements, and thus higher filter efficiencies
for reducing potential releases.
Changes to the operation of the containment spray system to
require operation during the recirculation mode are also mitigative
in nature. While the plant design basis has always included the
ability to implement containment spray during recirculation, this
license amendment now requires operation of containment spray in the
recirculation mode for dose mitigation. DCPP is designed and
licensed to operate using containment spray in the recirculation
mode. As such, operation of containment spray in the recirculation
mode has already been analyzed, evaluated, and is currently
controlled by Emergency Operating Procedures. Usage of recirculation
spray reduces the consequence of the postulated event. Likewise, the
additional shielding to the Control Room and the addition of a
[high-efficiency particulate air (HEPA)] filter to the [Technical
Support Center (TSC)] ventilation system reduces the consequences of
the postulated event to the Control Room and TSC personnel. Lowering
the limit for [Dose Equivalent XE-133 (DEX)] lowers potential
releases. By reclassifying a portion of the 40-inch Containment
Penetration Area Ventilation line and a portion of the 2-inch
gaseous radwaste system line to PG&E Design Class I, these lines
will be seismically qualified, thus assuring that post-LOCA release
points are the same as those used for determining [chi]/Q values.
The change to the decay time from 100 hours to 72 hours prior to
fuel movement is an input to the FHA. Although less decay will
result in higher released activity, the results of the FHA dose
consequence analysis remain within the dose acceptance criteria of
the event. Also, the radiation levels to an operator from a raised
fuel assembly may increase due to a lower decay time, however, any
exposure will continue to be maintained under 10 CFR 20 limits by
the plant Radiation Protection Program.
Plant-specific radiological analyses have been performed using
the AST methodology, assumption and inputs, as well as new [chi]/Q
values. The results of the dose consequences analyses demonstrate
that the regulatory acceptance criteria are met for each analyzed
event. Implementing the AST involves no facility equipment,
procedure, or process changes that could significantly affect the
radioactive material actually released during an event.
Subsequently, no conditions have been created that could
significantly increase the consequences of any of the events being
evaluated.
Based on the above discussion, 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 accident from any accident previously evaluated?
Response: No.
This license amendment does not alter or place any SSC in a
configuration outside its design or analysis limits and does not
create any new accident scenarios.
The AST methodology is not an accident initiator, as it is a
method used to estimate resulting postulated design basis accident
doses. The proposed TS changes reflect the plant configuration that
supports implementation of the new methodology and supports
reduction in dose consequences. DCPP is designed and licensed to
operate using containment spray in the recirculation mode. This
change will not affect any operational aspect of the system or any
other system, thus no new modes of operation are introduced by the
proposed change.
The function of the radiation monitors has not changed; only the
setpoint has changed as a result of an assessment of all potential
release pathways. The continued operation of containment spray and
the radiation monitor setpoint change do not create any new failure
modes, alter the nature of events postulated in the UFSAR, nor
introduce any unique precursor mechanism.
Requiring the 48-inch containment purge valves to be sealed
closed during operating MODES 1, 2, 3, and 4 does not introduce any
new accident precursor. This change only eliminates a potential
release path for radionuclides following a LOCA.
The proposed TS testing criteria for the auxiliary building
ventilation system charcoal filters and the proposed performance
criteria for steam generator tube integrity also cannot create an
accident, but results in requiring more efficient filtration of
potentially released iodine and less allowable primary-to-secondary
leakage. The proposed changes to the DEX activity limit, the TS
terminology, and the decay time of the fuel before movement are also
unrelated to accident initiators.
The only physical changes to the plant being made in support of
AST is the addition of Control Room shielding in an area previously
modified, the addition of a HEPA filter at the intake of the TSC
normal ventilation system, and the upgrade to the damper actuators,
pressure switches, and damper solenoid valves to support
reclassifying a portion of the Containment Penetration Area
Ventilation line to PG&E Design Class I. Both Control Room shielding
and HEPA filtration are mitigative in nature and do not have any
impact on plant operation or system response following an accident.
The Control Room modification for adding the shielding will meet
applicable loading limits, so the addition of the shielding cannot
initiate a failure. Upgrading damper actuators, pressure switches,
and damper solenoid valves involve replacing existing components
with components that are PG&E Design Class I. Therefore, the
addition of shielding, a HEPA filter, and upgrading components
cannot create a new or different kind of accident.
Since the function of the SSCs has not changed for AST
implementation, no new failure modes are created by this proposed
change. The AST change itself does not have the capability to
initiate accidents.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Implementing the AST is relevant only to calculated dose
consequences of potential design basis accidents evaluated in
Chapter 15 of the UFSAR. The changes proposed in this license
amendment involve the use of a new analysis methodology and related
regulatory acceptance criteria. New atmospheric dispersion factors,
which are based on site specific meteorological data, were
calculated in accordance with regulatory guidelines. The proposed
TS, TS Bases, and UFSAR changes reflect the plant configuration that
will support implementation of the new methodology and result in
operation in accordance with regulatory guidelines that support the
revisions to the radiological analyses of the limiting design basis
accidents. Conservative
[[Page 61488]]
methodologies, per the guidance of RG 1.183, have been used in
performing the accident analyses. The radiological consequences of
these accidents are all within the regulatory acceptance criteria
associated with the use of AST methodology.
The change to the minimum decay time prior to fuel movement
results in higher fission product releases after a FHA. However, the
results of the FHA dose consequence analysis remain within the dose
acceptance criteria of the event.
The proposed changes continue to ensure that the dose
consequences of design basis accidents at the exclusion area, low
population zone boundaries, in the TSC, and in the Control Room are
within the corresponding acceptance criteria presented in RG 1.183
and 10 CFR 50.67. The margin of safety for the radiological
consequences of these accidents is provided by meeting the
applicable regulatory limits, which are set at or below the 10 CFR
50.67 limits. An acceptable margin of safety is inherent in these
limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Michael T. Markley.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: June 30, 2015. A publicly-available
version is in ADAMS under Accession No. ML15181A470.
Description of amendment request: The amendment request proposes
changes to the Main Control Room Emergency Habitability System (VES)
configuration and equipment safety designation. Because, this proposed
change requires a departure from Tier 1 information in the Westinghouse
Advanced Passive 1000 Design Control Document (DCD), the licensee also
requested an exemption from the requirements of the Generic DCD Tier 1
in accordance with 10 CFR 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design functions of the VES for the main control room (MCR)
are to provide breathable air, maintain positive pressurization
relative to the outside, provide cooling of MCR equipment and
facilities, and provide passive air filtration within the MCR
boundary. The VES is designed to satisfy these functions for up to
72 hours following a design basis accident.
The proposed changes to the ASME [American Society of Mechanical
Engineers] safety classification of components, equipment
orientation and configuration, addition and deletion of components,
and correction to the number of emergency air storage tanks would
not adversely affect any design function. The proposed changes
maintain the design function of the VES with safety-related
equipment and system configuration consistent with the descriptions
in UFSAR [Updated Final Safety Analysis Report] Subsection 6.4.2.
The proposed changes do not affect the support or operation of
mechanical and fluid systems. There is no change to the response of
systems to postulated accident conditions. There is no change to the
predicted radioactive releases due to postulated accident
conditions. The plant response to previously evaluated accidents or
external events is not adversely affected, nor do the proposed
changes described create any new accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to revise the VES design related to the
ASME safety classification, equipment orientation and configuration,
addition and deletion of components, and correction to the number of
emergency air storage tanks maintains consistency with the design
function information in the USFAR. The proposed changes do not
create a new fault or sequence of events that could result in a
radioactive release. The proposed changes would not affect any
safety-related accident mitigating function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not affect the ability of the VES to
maintain the safety-related functions to the MCR. The VES continues
to meet the requirements for which it was designed and continues to
meet the regulations. No safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by the proposed changes,
and no margin of safety is reduced.
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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
III. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 17, 2015, as supplemented by
letters dated July 14, August 28, and September 3, 2015. Publicly-
available versions are in ADAMS under Accession Nos. ML15170A474,
ML15197A357, ML15243A044, and ML15246A638, respectively.
Brief description of amendment request: The amendment would modify
the technical specifications to define support systems needed in the
first 48 hours after a unit shutdown when steam generators are not
available for heat removal. The amendment would also make changes
consistent with Technical Specification Task Force Traveler-273-A,
Revision 2, to provide clarifications related to the requirements of
the Safety Function Determination Program.
Date of publication of individual notice in Federal Register:
September 15, 2015 (80 FR 55383).
[[Page 61489]]
Expiration date of individual notice: October 15, 2015 (public
comments); November 16, 2015 (hearing requests).
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: August 13, 2015. A publicly-available
version is in ADAMS under Accession No. ML15225A344.
Brief description of amendment request: To revise a current License
Condition (Section 2.F) regarding the Fire Protection Program and
propose a new License Condition regarding a fire protection
requirement.
Date of publication of individual notice in Federal Register:
September 4, 2015 (80 FR 53581).
Expiration date of individual notice: October 5, 2015 (public
comments); November 3, 2015 (hearing requests).
IV. 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. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: June 30, 2014, as supplemented
by letter dated June 8, 2015.
Brief description of amendments: The amendments revised the
Technical Specifications related to Technical Specification 3.5.2 by
reducing the allowed maximum Rated Thermal Power at which each unit can
operate when select High Pressure Injection system equipment is
inoperable.
Date of Issuance: September 24, 2015.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 395, 397 and 396. A publicly-available version is
in ADAMS under Accession No. ML15166A387; documents related to these
amendments are listed in the Safety Evaluation enclosure with the
amendments.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: September 16, 2014 (79
FR 55510). The supplement dated June 8, 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 September 24, 2015.
No significant hazards consideration comments received: No.
Duke Energy Progress, Docket No. 50-261, H. B. Robinson Steam Electric
Plant, Unit No. 2, Hartsville, South Carolina
Date of amendment request: February 10, 2014, as supplemented by
letters dated April 4, 2014, August 28, 2014, and September 4, 2015.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.3.1 for the Reactor Protection System
Instrumentation Turbine Trip function on Low Auto Stop Oil Pressure to
a Turbine Trip function on Low Electro-Hydraulic (EH) Fluid Oil
Pressure. The amendment revised the Allowable Value and Nominal Trip
Setpoint and revised the TS by applying additional testing requirements
listed in Technical Specification Task Force (TSTF) Traveler TSTF-493-
A, Revision 4, ``Clarify Application of Setpoint Methodologies for
Limiting Safety System Setting Functions,'' for Low EH Fluid Oil
Pressure trip.
Date of issuance: September 22, 2015.
Effective date: As of the date of issuance and shall be implemented
within 120 days of completion of the modification during Refueling
Outage 31 in fall of 2018.
Amendment No.: 243. A publicly-available version is in ADAMS under
Accession No. ML15040A073; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-23: Amendment revised
the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 22, 2014 (79 FR
42542). The supplemental letters dated August 28, 2014, and September
4, 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 as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 22, 2015.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: September 2, 2014, as supplemented by
letters dated April 23 and August 20, 2015.
Brief description of amendment: The amendment revised the
Surveillance Requirements (SRs) related to gas accumulation for the
emergency core cooling system and reactor core isolation cooling
system. The amendment also adds new SRs related to gas accumulation for
the residual heat removal and shutdown cooling systems. The NRC staff
has concluded that the Technical Specification (TS) changes are
consistent with NRC-approved Technical Specification Task Force (TSTF)
Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas
[[Page 61490]]
Accumulation,'' dated February 21, 2013, as part of the consolidated
line item improvement process. The TS Bases associated with these SRs
were also changed.
Date of issuance: September 21, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 188. A publicly-available version is in ADAMS under
Accession No. ML15195A061; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 6, 2015 (80 FR
522). The supplements dated April 23 and August 20, 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 as published in the Federal Register on January 6, 2015
(80 FR 522).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 21, 2015.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: November 3, 2014, as supplemented by
letter dated April 14, 2015.
Brief description of amendments: The amendments added new Limiting
Conditions for Operation (LCOs) 3.0.5 and 3.0.6 to the Applicability
section of the Technical Specifications (TSs). LCO 3.0.5 establishes an
allowance for restoring equipment to service under administrative
controls when the equipment has been removed from service or declared
inoperable to comply with TS Action requirements. LCO 3.0.6 provides
actions to be taken when the inoperability of a support system results
in the inoperability of the related supported systems. In addition, the
amendments added the Safety Function Determination Program to the
Administrative Controls section of the TSs. This program is intended to
ensure that a loss of safety function is detected and appropriate
actions are taken when LCO 3.0.6 is entered.
Date of issuance: September 15, 2015.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 219 (Unit 1) and 181 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML15218A501; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-39 and NPF-85:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: December 23, 2014 (79
FR 77046). The supplemental letter dated April 14, 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 as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 15, 2015.
No significant hazards consideration comments received: No.
National Institute of Standards and Technology (NIST), Docket No. 50-
184, Center for Neutron Research, National Bureau of Standards Test
Reactor (NBSR), Montgomery County, Maryland
Date of amendment request: June 23, 2014, as supplemented on August
20, 2014, February 26, 2015, and June 12, 2015.
Brief description of amendment: The amendment revised the NIST
NBSR's Technical Specifications Section 3.6 and Surveillance
Requirement 4.6, pertaining to the NIST reactor emergency power system,
which adds specifications and testing requirements for the new valve-
regulated lead acid batteries of the new uninterruptable power
supplies.
Date of issuance: September 10, 2015.
Effective date: As of the date of issuance.
Amendment No.: 10. A publicly-available version is in ADAMS under
Accession No. ML15237A146; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. TR-5: Amendment revised the Facility
Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 7, 2015 (80 FR
38760). The supplemental letters dated February 26, 2015, and June 12,
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 as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 10, 2015.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: April 29, 2015.
Brief description of amendment: The amendments revised the Updated
Final Safety Analysis Report (UFSAR) Table 15.6-17 to correct errors
introduced in UFSAR Revisions 16 and 17.
Date of issuance: September 22, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-207; Unit 2-195. A publicly-available
version is in ADAMS under Accession No. ML15209A641; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: July 21, 2015 (80 FR
43130).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 22, 2015.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: April 6, 2015, as supplemented by letter
dated July 15, 2015.
Brief description of amendment: The amendment revised the Technical
Specifications by modifying the acceptance criteria for the emergency
diesel generator steady-state frequency range in associated
surveillance requirements.
Date of issuance: September 17, 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.: 102. A publicly-available version is in ADAMS under
Accession No. ML15230A155;
[[Page 61491]]
documents related to this amendment are listed in the Safety Evaluation
enclosed with the amendment.
Facility Operating License No. NFP-90: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 26, 2015 (80 FR
30103). The supplemental letter dated July 15, 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 as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 17, 2015.
No significant hazards consideration determination comments
received: No.
V. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses and Final Determination of No Significant Hazards
Consideration and Opportunity for a Hearing (Exigent Public
Announcement or Emergency Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual notice of
consideration of issuance of amendment, proposed no significant hazards
consideration determination, and opportunity for a hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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 51.12(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
application for amendment, (2) the amendment to Facility Operating
License or Combined License, as applicable, 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.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC's Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR's Reference staff
at 1-800-397-4209, 301-415-4737, or by email to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the
[[Page 61492]]
requestor's/petitioner's interest. The petition must also identify the
specific contentions which the requestor/petitioner seeks to have
litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the 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 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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
Arizona Public Service Company, Docket No. 50-529, Palo Verde Nuclear
Generating Station, Unit 2, Maricopa County, Arizona
Date of amendment request: September 4, 2015, as supplemented by
letter dated September 15, 2015.
Description of amendment request: The amendment added a Note to
Technical Specification Surveillance Requirement (SR) 3.1.5.3, Control
Element Assembly (CEA) freedom of movement surveillance, such that Unit
2, CEA 88 may be excluded from the remaining quarterly performance of
the SR in Unit 2, Cycle 19 due to a degraded upper gripper coil. The
amendment allows the licensee to delay exercising CEA 88 until after
repairs can be made during the upcoming fall 2015 outage.
Date of issuance: September 25, 2015.
Effective date: This license amendment is effective as of the date
of issuance and shall be implemented prior to the SR 3.1.5.3
performance due date for CEA 88 in Unit 2, Cycle 19.
Amendment No.: 196. A publicly-available version is in ADAMS under
Accession No. ML15266A005; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-51: Amendment revised
the Operating License and Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Public notice of the proposed amendment was
published in the Arizona Republic, located in Phoenix, Arizona, from
September 21 through September 22, 2015. The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments were received. The supplemental letter dated
September 15, 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 NSHC
determination as published in the Arizona Republic.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a Safety Evaluation dated September 25, 2015.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Dated at Rockville, Maryland, this 1st day of October 2015.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-25860 Filed 10-9-15; 8:45 am]
BILLING CODE 7590-01-P