Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 61476-61492 [2015-25860]

Agencies

• NUCLEAR REGULATORY COMMISSION

[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 https://www.regulations.gov and search for Docket ID NRC-2015-0236. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov.
     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: Janet.Burkhardt@nrc.gov.

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 https://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 https://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 pdr.resource@nrc.gov. 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 https://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 https://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, https://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 pdr.resource@nrc.gov. 
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