28456 Federal Register / Vol. 83, No. 118 / Tuesday, June 19, 2018 / Notices 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2018–0114] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses and Combined Licenses 

Involving No Significant Hazards 

Considerations 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Biweekly notice. 

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 May 22, 

2018, to June 4, 2018. The last biweekly 

notice was published on June 5, 2018. 

DATES: Comments must be filed by July 

19, 2018. A request for a hearing must 

be filed by August 20, 2018. 

ADDRESSES: You may submit comments 

by any of the following methods (unless 

this document describes a different 

method for submitting comments on a 

specific subject): 

• Federal Rulemaking website: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2018–0114. Address 

questions about NRC dockets to Jennifer 

Borges; telephone: 301–287–9127; 

email: Jennifer.Borges@nrc.gov. For 

technical questions, contact the 

individual listed in the FOR FURTHER

INFORMATION CONTACT section of this 

document. 

• Mail comments to: May Ma, Office 

of Administration, Mail Stop: TWFN–7– 

A60M, 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–2018– 

0114, facility name, unit number(s), 

plant docket number, application date, 

and subject when contacting the NRC 

about the availability of information for 

this action. You may obtain publiclyavailable information related to this 

action by any of the following methods: 

• Federal Rulemaking website: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2018–0114. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publiclyavailable documents online in the 

ADAMS Public Documents collection at 

http://www.nrc.gov/reading-rm/ 

adams.html. To begin the search, select 

‘‘ADAMS Public Documents’’ and then 

select ‘‘Begin Web-based ADAMS 

Search.’’ For problems with ADAMS, 

please contact the NRC’s Public 

Document Room (PDR) reference staff at 

1–800–397–4209, 301–415–4737, or by 

email to 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 this document 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

B. Submitting Comments 

Please include Docket ID NRC–2018– 

0114, facility name, unit number(s), 

plant docket number, application date, 

and subject in your comment 

submission. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 

disclosed in your comment submission. 

The NRC will post all comment 

submissions at http://

www.regulations.gov as well as enter the 

comment submissions into ADAMS. 

The NRC does not routinely edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information that 

they do not want to be publicly 

disclosed in their comment submission. 

Your request should state that the NRC 

does not routinely edit comment 

submissions to remove such information 

before making the comment 

submissions available to the public or 

entering the comment into ADAMS. 

II. Notice of Consideration of Issuance 

of Amendments to Facility Operating 

Licenses and Combined Licenses and 

Proposed No Significant Hazards 

Consideration Determination 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

section 50.92 of title 10 of the Code of 

Federal Regulations (10 CFR), this 

means that operation of the facility in 

accordance with the proposed 

amendment would not (1) involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated, or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period if circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. If 

the Commission takes action prior to the 

expiration of either the comment period 

or the notice period, it will publish in 

the Federal Register a notice of 

issuance. If the Commission makes a 

final no significant hazards 

consideration determination, any 

hearing will take place after issuance. 

The Commission expects that the need 

to take this action will occur very 

infrequently. 


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A. Opportunity To Request a Hearing 

and Petition for Leave To Intervene 

Within 60 days after the date of 

publication of this notice, any persons 

(petitioner) whose interest may be 

affected by this action may file a request 

for a hearing and petition for leave to 

intervene (petition) with respect to the 

action. Petitions shall be filed in 

accordance with the Commission’s 

‘‘Agency Rules of Practice and 

Procedure’’ in 10 CFR part 2. Interested 

persons should consult a current copy 

of 10 CFR 2.309. The NRC’s regulations 

are accessible electronically from the 

NRC Library on the NRC’s website at 

http://www.nrc.gov/reading-rm/doccollections/cfr/. Alternatively, a copy of 

the regulations is available at the NRC’s 

Public Document Room, located at One 

White Flint North, Room O1–F21, 11555 

Rockville Pike (First Floor), Rockville, 

Maryland 20852. If a petition is filed, 

the Commission or a presiding officer 

will rule on the petition and, if 

appropriate, a notice of a hearing will be 

issued. 

As required by 10 CFR 2.309(d), the 

petition should specifically explain the 

reasons why intervention should be 

permitted with particular reference to 

the following general requirements for 

standing: (1) The name, address, and 

telephone number of the petitioner; (2) 

the nature of the petitioner’s right under 

the Act to be made a party to the 

proceeding; (3) the nature and extent of 

the petitioner’s property, financial, or 

other interest in the proceeding; and (4) 

the possible effect of any decision or 

order which may be entered in the 

proceeding on the petitioner’s interest. 

In accordance with 10 CFR 2.309(f), 

the petition must also set forth the 

specific contentions which the 

petitioner seeks to have litigated in the 

proceeding. Each contention must 

consist of a specific statement of the 

issue of law or fact to be raised or 

controverted. In addition, the petitioner 

must provide a brief explanation of the 

bases for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to the specific 

sources and documents on which the 

petitioner intends to rely to support its 

position on the issue. The petition must 

include sufficient information to show 

that a genuine dispute exists with the 

applicant or licensee on a material issue 

of law or fact. Contentions must be 

limited to matters within the scope of 

the proceeding. The contention must be 

one which, if proven, would entitle the 

petitioner to relief. A petitioner who 

fails to satisfy the requirements at 10 

CFR 2.309(f) with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene. Parties have the opportunity 

to participate fully in the conduct of the 

hearing with respect to resolution of 

that party’s admitted contentions, 

including the opportunity to present 

evidence, consistent with the NRC’s 

regulations, policies, and procedures. 

Petitions must be filed no later than 

60 days from the date of publication of 

this notice. Petitions and motions for 

leave to file new or amended 

contentions that are filed after the 

deadline will not be entertained absent 

a determination by the presiding officer 

that the filing demonstrates good cause 

by satisfying the three factors in 10 CFR 

2.309(c)(1)(i) through (iii). The petition 

must be filed in accordance with the 

filing instructions in the ‘‘Electronic 

Submissions (E-Filing)’’ section of this 

document. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to 

establish when the hearing is held. If the 

final determination is that the 

amendment request involves no 

significant hazards consideration, the 

Commission may issue the amendment 

and make it immediately effective, 

notwithstanding the request for a 

hearing. Any hearing would take place 

after issuance of the amendment. If the 

final determination is that the 

amendment request involves a 

significant hazards consideration, then 

any hearing held would take place 

before the issuance of the amendment 

unless the Commission finds an 

imminent danger to the health or safety 

of the public, in which case it will issue 

an appropriate order or rule under 10 

CFR part 2. 

A State, local governmental body, 

Federally-recognized Indian Tribe, or 

agency thereof, may submit a petition to 

the Commission to participate as a party 

under 10 CFR 2.309(h)(1). The petition 

should state the nature and extent of the 

petitioner’s interest in the proceeding. 

The petition should be submitted to the 

Commission no later than 60 days from 

the date of publication of this notice. 

The petition must be filed in accordance 

with the filing instructions in the 

‘‘Electronic Submissions (E-Filing)’’ 

section of this document, and should 

meet the requirements for petitions set 

forth in this section, except that under 

10 CFR 2.309(h)(2) a State, local 

governmental body, or Federallyrecognized Indian Tribe, or agency 

thereof does not need to address the 

standing requirements in 10 CFR 

2.309(d) if the facility is located within 

its boundaries. Alternatively, a State, 

local governmental body, Federallyrecognized Indian Tribe, or agency 

thereof may participate as a non-party 

under 10 CFR 2.315(c). 

If a hearing is granted, any person 

who is not a party to the proceeding and 

is not affiliated with or represented by 

a party may, at the discretion of the 

presiding officer, be permitted to make 

a limited appearance pursuant to the 

provisions of 10 CFR 2.315(a). A person 

making a limited appearance may make 

an oral or written statement of his or her 

position on the issues but may not 

otherwise participate in the proceeding. 

A limited appearance may be made at 

any session of the hearing or at any 

prehearing conference, subject to the 

limits and conditions as may be 

imposed by the presiding officer. Details 

regarding the opportunity to make a 

limited appearance will be provided by 

the presiding officer if such sessions are 

scheduled. 

B. Electronic Submissions (E-Filing) 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing and petition for 

leave to intervene (petition), any motion 

or other document filed in the 

proceeding prior to the submission of a 

request for hearing or petition to 

intervene, and documents filed by 

interested governmental entities that 

request to participate under 10 CFR 

2.315(c), must be filed in accordance 

with the NRC’s E-Filing rule (72 FR 

49139; August 28, 2007, as amended at 

77 FR 46562; August 3, 2012). The EFiling process requires participants to 

submit and serve all adjudicatory 

documents over the internet, or in some 

cases to mail copies on electronic 

storage media. Detailed guidance on 

making electronic submissions may be 

found in the Guidance for Electronic 

Submissions to the NRC and on the NRC 

website at http://www.nrc.gov/site-help/ 

e-submittals.html. Participants may not 

submit paper copies of their filings 

unless they seek an exemption in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least 10 

days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by email at 


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hearing.docket@nrc.gov, or by telephone 

at 301–415–1677, to (1) request a digital 

identification (ID) certificate, which 

allows the participant (or its counsel or 

representative) to digitally sign 

submissions and access the E-Filing 

system for any proceeding in which it 

is participating; and (2) advise the 

Secretary that the participant will be 

submitting a petition or other 

adjudicatory document (even in 

instances in which the participant, or its 

counsel or representative, already holds 

an NRC-issued digital ID certificate). 

Based upon this information, the 

Secretary will establish an electronic 

docket for the hearing in this proceeding 

if the Secretary has not already 

established an electronic docket. 

Information about applying for a 

digital ID certificate is available on the 

NRC’s public website at http://

www.nrc.gov/site-help/e-submittals/ 

getting-started.html. Once a participant 

has obtained a digital ID certificate and 

a docket has been created, the 

participant can then submit 

adjudicatory documents. Submissions 

must be in Portable Document Format 

(PDF). Additional guidance on PDF 

submissions is available on the NRC’s 

public website at http://www.nrc.gov/ 

site-help/electronic-sub-ref-mat.html. A 

filing is considered complete at the time 

the document is submitted through the 

NRC’s E-Filing system. To be timely, an 

electronic filing must be submitted to 

the E-Filing system no later than 11:59 

p.m. Eastern Time on the due date. 

Upon receipt of a transmission, the EFiling system time-stamps the document 

and sends the submitter an email notice 

confirming receipt of the document. The 

E-Filing system also distributes an email 

notice that provides access to the 

document to the NRC’s Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the document on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before adjudicatory 

documents are filed so that they can 

obtain access to the documents via the 

E-Filing system. 

A person filing electronically using 

the NRC’s adjudicatory E-Filing system 

may seek assistance by contacting the 

NRC’s Electronic Filing Help Desk 

through the ‘‘Contact Us’’ link located 

on the NRC’s public website at http://

www.nrc.gov/site-help/esubmittals.html, by email to 

MSHD.Resource@nrc.gov, or by a tollfree call at 1–866–672–7640. The NRC 

Electronic Filing Help Desk is available 

between 9 a.m. and 6 p.m., Eastern 

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing stating why there is good cause for 

not filing electronically and requesting 

authorization to continue to submit 

documents in paper format. Such filings 

must be submitted by: (1) First class 

mail addressed to the Office of the 

Secretary of the Commission, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Rulemaking and Adjudications Staff; or 

(2) courier, express mail, or expedited 

delivery service to the Office of the 

Secretary, 11555 Rockville Pike, 

Rockville, Maryland 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing adjudicatory 

documents in this manner are 

responsible for serving the document on 

all other participants. Filing is 

considered complete by first-class mail 

as of the time of deposit in the mail, or 

by courier, express mail, or expedited 

delivery service upon depositing the 

document with the provider of the 

service. A presiding officer, having 

granted an exemption request from 

using E-Filing, may require a participant 

or party to use E-Filing if the presiding 

officer subsequently determines that the 

reason for granting the exemption from 

use of E-Filing no longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in the NRC’s 

electronic hearing docket which is 

available to the public at https://

adams.nrc.gov/ehd, unless excluded 

pursuant to an order of the Commission 

or the presiding officer. If you do not 

have an NRC-issued digital ID certificate 

as described above, click cancel when 

the link requests certificates and you 

will be automatically directed to the 

NRC’s electronic hearing dockets where 

you will be able to access any publicly 

available documents in a particular 

hearing docket. Participants are 

requested not to include personal 

privacy information, such as social 

security numbers, home addresses, or 

personal phone numbers in their filings, 

unless an NRC regulation or other law 

requires submission of such 

information. For example, in some 

instances, individuals provide home 

addresses in order to demonstrate 

proximity to a facility or site. With 

respect to copyrighted works, except for 

limited excerpts that serve the purpose 

of the adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

For further details with respect to 

these license amendment applications, 

see the application for amendment 

which is available for public inspection 

in ADAMS and at the NRC’s PDR. For 

additional direction on accessing 

information related to this document, 

see the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

Duke Energy Progress, LLC, Docket Nos. 

50–325 and 50–324, Brunswick Steam 

Electric Plant, Unit Nos. 1 and 2, 

Brunswick County, North Carolina 

Date of amendment request: January 

23, 2018. A publicly-available version is 

in ADAMS under Accession No. 

ML18023A896. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specification 3.6.4.1, 

‘‘Secondary Containment,’’ Surveillance 

Requirement (SR) 3.6.4.1.2, for 

Brunswick Steam Electric Plant, Units 1 

and 2. The proposed changes are based 

on Technical Specifications Task Force 

(TSTF) Traveler TSTF–551, Revision 3, 

‘‘Revise Secondary Containment 

Surveillance Requirements’’ (ADAMS 

Accession No. ML16277A226). 

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 addresses conditions 

during which Secondary Containment SR 

3.6.4.1.2 is not met. The Secondary 

Containment is not an initiator of any 

accident previously evaluated. As a result, 

the probability of any accident previously 

evaluated is not increased. The consequences 

of an accident previously evaluated while 

utilizing the proposed change is no different 

than the consequences of an accident while 

utilizing the existing eight hour Completion 

Time for an inoperable Secondary 

Containment. As a result, the consequences 

of an 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 amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not alter the 

protection system design, create new failure 


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modes, or change any modes of operation. 

The proposed change does not involve a 

physical alteration of the plant; and no new 

or different kind of equipment will be 

installed. Consequently, there are no new 

initiators that could result in a new or 

different kind of accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change addresses conditions 

during which Secondary Containment SR 

3.6.4.1.2 is not met. The allowance for both 

an inner and outer Secondary Containment 

door to be open simultaneously for entry and 

exit does not affect the safety function of the 

Secondary Containment as the doors are 

promptly closed after entry or exit, thereby 

restoring the Secondary Containment 

boundary. 

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: Kathryn B. 

Nolan, Deputy General Counsel, 550 

South Tryon Street, Mail Code DEC45A, 

Charlotte, NC 28202. 

NRC Acting Branch Chief: Brian W. 

Tindell. 

Duke Energy Progress, LLC, Docket Nos. 

50–325 and 50–324, Brunswick Steam 

Electric Plant, Unit Nos. 1 and 2, 

Brunswick County, North Carolina 

Date of amendment request: January 

23, 2018. A publicly-available version is 

in ADAMS under Accession No. 

ML18023A899. 

Description of amendment request: 

The amendments would revise the 

Technical Specifications to adopt 

Technical Specifications Task Force 

(TSTF) Traveler TSTF–208, Revision 0, 

‘‘Extension of Time to Reach Mode 2 in 

LCO [Limiting Condition for Operation] 

3.0.3.’’ 

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 time frame to take response action in 

accordance with LCO 3.0.3 is not an 

initiating condition for any accident 

previously evaluated. The proposed change 

does not authorize the addition of any new 

plant equipment or systems, nor does it alter 

the assumptions of any accident analyses. 

The small increase in the time allowed to 

reach Mode 2 would not place the plant in 

any significantly increased probability of an 

accident occurring. The unit would already 

be preparing for a plant shutdown condition 

because of the 1 hour requirement to initiate 

shutdown actions. There is no change in the 

time period to reach Mode 3. The Mode 3 

Condition is the point at which the plant 

reactor core is no longer critical (i.e., Hot 

Shutdown). 

Therefore, since there is no change to the 

time period to reach the Hot Shutdown 

condition, the small change in the time to 

reach Mode 2 status 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 allowed time 

to reach Mode 2 in LCO 3.0.3 does not 

require any modification to the plant or 

change equipment operation. The proposed 

change will not introduce failure modes that 

could result in a new accident, and the 

change does not alter assumptions made in 

the safety analysis. The proposed change will 

not alter the design configuration, or method 

of operation of plant equipment beyond its 

normal functional capabilities. The proposed 

change does not create any new credible 

failure mechanisms, malfunctions, or 

accident initiators. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from those that have been 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change to the allowed time 

to reach Mode 2 in LCO 3.0.3 does not alter 

or exceed a design basis or safety limit. There 

is no change being made to safety analysis 

assumptions or the safety limits that would 

adversely affect plant safety as a result of the 

proposed change. Margins of safety are 

unaffected by the proposed change and the 

applicable requirements of 10 CFR 

50.36(c)(2)(ii) and 10 CFR 50, Appendix A 

will continue to be met. 

Therefore, the proposed change does not 

involve any 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: Kathryn B. 

Nolan, Deputy General Counsel, 550 

South Tryon Street, M/C DEC45A, 

Charlotte, NC 28202. 

NRC Acting Branch Chief: Brian W. 

Tindell. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit No. 3 (Waterford 3), St. Charles 

Parish, Louisiana 

Date of amendment request: March 8, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18068A705. 

Description of amendment request: 

The amendment would update Section 

15.4.3.1 of the Updated Final Safety 

Analysis Report for Waterford 3, which 

describes the dose consequence of the 

worst undetectable single fuel assembly 

misload. The updated analysis would 

reflect the use of Next Generation Fuel 

and integrated fuel burnable absorbers. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the fuel 

assembly misload event analysis. The 

analysis of the fuel assembly misload event 

showed that the total number of failed fuel 

rods is less than other Waterford 3 Condition 

3 events that have already been demonstrated 

to meet the 10 CFR 50.67 acceptance criteria. 

For Waterford 3, the Excess Load with Loss 

of Alternating Current (LOAC) has this same 

release and fuel failure that has been shown 

to meet the offsite dose requirements. Since 

the worst undetectable misload has a fuel 

failure less than the excess load with LOAC 

event, the fuel assembly misload event is 

consistent with the Standard Review Plan 

15.4.7 and meets the 10 CFR 50.67 

requirements. 

This change is only analyzing the 

consequences of the fuel assembly misload 

event and no changes are being made that 

would impact the probability of the event 

occurring. 

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 fuel 

assembly misload event analysis. The 

proposed change does not involve a physical 

alteration of the plant (no new or different 

type of equipment will be installed) or a 

change in the methods governing plant 

operations. The proposed change will not 

introduce new failure modes or effects and 

will not, in the absence of other unrelated 

failures, lead to an accident whose 


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consequences exceed the consequences of 

accidents previously analyzed. 

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 fuel 

assembly misload event analysis. The worst 

undetectable misloads have fuel failure less 

than the excess load with the Excess Load 

with Loss of Alternating Current (LOAC) 

event; the fuel assembly misload event meets 

the 10 CFR 50.67 criteria and is consistent 

with the Standard Review Plan Section 

15.4.7 guidance. The new analysis shows 

more adverse consequences than were shown 

in previous fuel assembly misload event 

analyses, but remains within the regulatory 

acceptance limits. Since the event remains 

within the 10 CFR 50.67 requirements and is 

bounded by the excess load with LOAC 

event, this is not a significant reduction in 

margin. 

Therefore, 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: Anna Vinson 

Jones, Senior Counsel, Entergy Services, 

Inc., 101 Constitution Avenue NW, 

Suite 200 East, Washington, DC 20001. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Unit 

Nos. 2 and 3, Grundy County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Unit Nos. 1 and 2, 

LaSalle County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Unit Nos. 

1 and 2, Rock Island County, Illinois 

Date of amendment request: April 25, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18116A133. 

Description of amendment request: 

The amendments would revise the 

technical specification (TS) 

requirements for inoperable snubbers 

for each facility. The amendments 

would also make other administrative 

changes to the TSs. 

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 for each site, 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 allows a delay time 

before declaring supported Technical 

Specification (TS) systems inoperable when 

the associated snubber(s) cannot perform its 

required safety function. Entrance into 

Actions or delaying entrance into Actions is 

not an initiator of any accident previously 

evaluated. Consequently, the probability of 

an accident previously evaluated is not 

significantly increased. The consequences of 

an accident while relying on the delay time 

allowed before declaring a TS supported 

system inoperable and taking its Conditions 

and Required Actions are no different than 

the consequences of an accident under the 

same plant conditions while relying on the 

existing TS supported system Conditions and 

Required Actions. Therefore, the 

consequences of an accident previously 

evaluated are not significantly increased by 

this change. 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 allows a delay time 

before declaring supported TS systems 

inoperable when the associated snubber(s) 

cannot perform its required safety function. 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. 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 allows a delay time 

before declaring supported TS systems 

inoperable when the associated snubber(s) 

cannot perform its required safety function. 

The proposed change restores an allowance 

in the pre-Improved Standard Technical 

Specifications (ISTS) conversion TS that was 

unintentionally eliminated by the 

conversion. The pre-ISTS TS were 

considered to provide an adequate margin of 

safety for plant operation, as does the postISTS conversion TS. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis for each site and, 

based on this review, it appears that the 

three standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Tamra Domeyer, 

Associate General Counsel, Exelon 

Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: David J. Wrona. 

FirstEnergy Nuclear Operating 

Company, Docket No. 50–440, Perry 

Nuclear Power Plant, Unit No. 1, Lake 

County, Ohio 

Date of amendment request: March 7, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18066A648. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification 5.5.12, 

‘‘Primary Containment Leakage Rate 

Testing Program,’’ to follow guidance 

developed by the Nuclear Energy 

Institute (NEI) in topical report NEI 94– 

01, ‘‘Industry Guideline for 

Implementing Performance-Based 

Option of 10 CFR part 50, Appendix J,’’ 

Revision 3–A, dated July 2012, with the 

conditions and limitations specified in 

NEI 94–01, Revision 2–A, dated October 

2008. The proposed license amendment 

would also revise Technical 

Specification 5.5.12 by deleting two of 

the four listed exceptions to program 

guidelines. 

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 test interval extensions do 

not involve either a physical change to the 

plant or a change in the way 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. Therefore, the 

proposed extensions do not involve a 

significant increase in the probability of an 

accident previously evaluated. 

The effect resulting from changing the 

Type A test frequency to 1 per 15 years, 

measured as an increase to the total 

integrated plant risk for those accident 

sequences influenced by Type A testing, is 


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0.0318 person-rem/year. EPRI [Electric Power 

Research Institute] Report No. 1009325, 

Revision 2–A, states that a very small 

population dose is defined as an increase of 

less than or equal to 1.0 person-rem per year 

or less than or equal to 1 percent of the total 

population dose, whichever is less restrictive 

for the risk impact assessment of the 

extended integrated leak rate test intervals. 

The results of the risk assessment calculation 

for the Type A test extension meet these 

criteria. The risk impact for the integrated 

leak rate test extension when compared to 

other severe accident risks is negligible. 

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 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 for Mechanical Engineers 

Boiler and Pressure Vessel Code (ASME 

Code)], Section XI, and Technical 

Specification requirements serve to provide a 

high degree of assurance that the 

containment would not degrade in a manner 

that is detectable only by a Type A test. 

Based on the above, the proposed test 

interval extensions do not significantly 

increase the consequences of an accident 

previously evaluated. 

The proposed amendment also deletes two 

previously granted exceptions to Primary 

Containment Leakage Rate Testing Program 

guidelines. The exception regarding the 

performance of a Type A test no later than 

a specified date would be deleted as this 

Type A test has already been performed. 

Additionally, the exception to use the 

corrections to NEI 94–01, Revision 0, would 

be deleted as those corrections would no 

longer be in use. These changes to the 

exceptions in Technical Specification 5.5.12 

are administrative in nature and do not affect 

the probability or consequences of an 

accident previously evaluated. 

Therefore, the proposed changes do 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. 

Containment Type A and Type C testing 

requirements periodically demonstrate the 

integrity of the containment and exist to 

ensure the plant’s ability to mitigate the 

consequences of an accident. These tests do 

not involve any accident precursors or 

initiators. 

The proposed change does not involve a 

physical modification to the plant (that is, no 

new or different type of equipment will be 

installed) nor does it alter the design, 

configuration, or change the manner in 

which the plant is operated or controlled 

beyond the standard functional capabilities 

of the equipment. 

The proposed amendment also deletes two 

previously granted exceptions. The exception 

regarding the performance of a Type A test 

no later than a specified date would be 

deleted as this Type A test has already been 

performed. Additionally, the exception to use 

the corrections to NEI 94–01, Revision 0, 

would be deleted as those corrections would 

no longer be in use. These changes to the 

exceptions in Technical Specification 5.5.12 

are administrative in nature and do not create 

the possibility of a new or different kind of 

accident from any previously evaluated. 

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 license amendment does not 

alter the way safety limits, limiting safety 

system set points, or limiting conditions for 

operation are determined. The specific 

requirements and conditions of the Technical 

Specification Primary Containment Leakage 

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 Technical Specifications 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 amendment, since they are not 

affected by implementation of a performancebased containment testing program. This 

ensures that the margin of safety in the plant 

safety analysis is maintained. 

The proposed amendment also deletes two 

previously granted exceptions. The exception 

regarding the performance of a Type A test 

no later than a specified date would be 

deleted as this Type A test has already been 

performed. Additionally, the exception to use 

the corrections to NEI 94–01, Revision 0, 

would be deleted as those corrections would 

no longer be in use. These changes to the 

exceptions in Technical Specification 5.5.12 

are administrative in nature and do not 

involve a significant reduction in a margin of 

safety. 

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 W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop A–GO–15, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: David J. Wrona. 

NextEra Energy, Point Beach, LLC, 

Docket Nos. 50–266 and 50–301, Point 

Beach Nuclear Plant (PBNP), Unit Nos. 

1 and 2, Town of Two Creeks, 

Manitowoc County, Wisconsin 

Date of amendment request: March 

30, 2018. A publicly-available version is 

in ADAMS under Accession No. 

ML18092A239. 

Description of amendment request: 

The amendments would revise 

Technical Specification (TS) 5.5.15, 

‘‘Containment Leakage Rate Testing 

Program,’’ to require a program in 

accordance with Nuclear Energy 

Institute (NEI) topical report NEI 94–01, 

Revision 3–A, ‘‘Industry Guideline for 

Implementing Performance-Based 

Option of 10 CFR part 50, Appendix J.’’ 

This proposed change will allow 

extension of the Type A test interval up 

to one test in 15 years and extension of 

the Type C test interval up to 75 

months, based on acceptable 

performance history as defined in NEI 

94–01, Revision 3–A. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment adopts the NRCaccepted guidelines of NEI 94–01, Revision 

3–A, ‘‘Industry Guideline for Implementing 

Performance-Based Option of 10 CFR part 50, 

Appendix J,’’ for development of the PBNP 

performance-based containment testing 

program. NEI 94–01 allows, based on risk 

and performance, an extension of Type A and 

Type C containment leak test intervals. 

Implementation of these guidelines continues 

to provide adequate assurance that during 

design basis accidents, the primary 

containment and its components will limit 

leakage rates to less than the values assumed 

in the plant safety analyses. 

The findings of the PBNP risk assessment 

confirm the general findings of previous 

studies that the risk impact with extending 

the containment leak rate is small. Per the 

guidance provided in Regulatory Guide 

1.174, an extension of the leak test interval 

in accordance with NEI 94–01, Revision 3– 

A results in an estimated change within, the 

very small change region. 

Since the change is implementing a 

performance-based containment testing 

program, the proposed amendment 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 requirement 

for containment leakage rate acceptance will 

not be changed by this amendment. 


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Therefore, the containment will continue to 

perform its design function as a barrier to 

fission product releases. 

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 implement a 

performance-based containment testing 

program, associated with integrated leakage 

rate test frequency, does not change the 

design or operation of structures, systems, or 

components of the plant. 

The proposed change would continue to 

ensure containment integrity and would 

ensure operation within the bounds of 

existing accident analyses. There are no 

accident initiators created or affected by this 

change. Therefore, the proposed change will 

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. 

Margin of safety is related to confidence in 

the ability of the fission product barriers (fuel 

cladding, reactor coolant system, and 

primary containment) to perform their design 

functions during and following postulated 

accidents. The proposed change to 

implement a performance-based containment 

testing program, associated with integrated 

leakage rate test and local leak rate testing 

frequency, does not affect plant operations, 

design functions, or any analysis that verifies 

the capability of a structure, system, or 

component of the plant to perform a design 

function. In addition, this change does not 

affect safety limits, limiting safety system 

setpoints, or limiting conditions for 

operation. 

The specific requirements and conditions 

of the TS Containment Leakage Rate Testing 

Program exist to ensure that the degree of 

containment structural integrity and leaktightness that is considered in the plant 

safety analysis is maintained. The overall 

containment leak rate limit specified by TS 

is maintained. This 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 implementation of a performancebased containment testing program. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William Blair, 

Managing Attorney—Nuclear, Florida 

Power & Light Company, P.O. Box 

14000, 700 Universe Boulevard, Juno 

Beach, FL 33408–0420. 

NRC Branch Chief: David J. Wrona. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station (HCGS), 

Salem County, New Jersey 

Date of amendment request: April 13, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18103A218. 

Description of amendment request: 

The amendment would revise Technical 

Specification (TS) 3.8.3.1, 

‘‘Distribution—Operating,’’ to increase 

the alternating current (AC) inverters 

allowed outage time (AOT) from 24 

hours to 7 days. The proposed change 

is based on application of the HCGS 

probabilistic risk assessment (PRA) in 

support of a risk-informed extension, 

and on additional considerations and 

compensatory actions. 

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 TS amendment does not 

affect the design of the AC inverters, the 

operational characteristics or function of the 

inverters, the interfaces between the inverters 

and other plant systems, or the reliability of 

the inverters. An inoperable AC inverter is 

not considered an initiator of an analyzed 

event. In addition, TS Actions and the 

associated Allowed Outage Times are not 

initiators of previously evaluated accidents. 

Extending the Allowed Outage Time for an 

inoperable AC inverter would not have a 

significant impact on the frequency of 

occurrence of an accident previously 

evaluated. The proposed amendment will not 

result in modifications to plant activities 

associated with inverter maintenance, but 

rather, provides operational flexibility by 

allowing additional time to perform inverter 

troubleshooting, corrective maintenance, and 

post-maintenance testing on-line. 

The proposed extension of the Completion 

Time for an inoperable AC inverter will not 

significantly affect the capability of the 

inverters to perform their safety function, 

which is to ensure an uninterruptible supply 

of 120-volt AC electrical power to the 

associated power distribution subsystems. 

An evaluation, using PRA methods, 

confirmed that the increase in plant risk 

associated with implementation of the 

proposed Allowed Outage Time extension is 

consistent with the NRC’s Safety Goal Policy 

Statement, as further described in RG 

[Regulatory Guide] 1.174 and RG 1.177. In 

addition, a deterministic evaluation 

concluded that plant defense-in-depth 

philosophy will be maintained with the 

proposed Allowed Outage Time extension. 

There will be no impact on the source term 

or pathways assumed in accidents previously 

evaluated. No analysis assumptions will be 

changed and there will be no adverse effects 

on onsite or offsite doses as the result of an 

accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment does not involve 

physical alteration of the HCGS. No new 

equipment is being introduced, and installed 

equipment is not being operated in a new or 

different manner. There is no change being 

made to the parameters with in which the 

HCGS is operated. There are no setpoints at 

which protective or mitigating actions are 

initiated that are affected by this proposed 

action. The use of the alternate Class 1E 

power source for the AC distribution panel 

is consistent with the HCGS plant design. 

The change does not alter assumptions made 

in the safety analysis. This proposed action 

will not alter the manner in which 

equipment operation is initiated, nor will the 

functional demands on credited equipment 

be changed. No alteration is proposed to the 

procedures that ensure the HCGS remains 

with in analyzed limits, and no change is 

being made to procedures relied upon to 

respond to an off-normal event. As such, no 

new failure modes are being introduced. 

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. 

Margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their design 

functions during and following an accident. 

These barriers include the fuel cladding, the 

reactor coolant system, and the containment 

system. The proposed change, which would 

increase the AOT from 24 hours to 7 days for 

one inoperable inverter, does not exceed or 

alter a setpoint, design basis or safety limit. 

Therefore, the proposed amendment 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. 


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Attorney for licensee: Jeffrie J. Keenan, 

PSEG Nuclear LLC—N21, P.O. Box 236, 

Hancocks Bridge, NJ 08038. 

NRC Branch Chief: James G. Danna. 

Southern Nuclear Operating Company, 

Docket Nos. 52–025 and 52–026, Vogtle 

Electric Generating Plant, Unit Nos. 3 

and 4, Burke County, Georgia 

Date of amendment request: April 26, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18116A138. 

Description of amendment request: 

The requested amendment proposes 

changes to combined license (COL) 

Appendix C, with corresponding 

changes to the associated plant-specific 

Tier 1 information, and involves 

associated Tier 2 information in the 

Updated Final Safety Analysis Report 

(UFSAR) (which includes the plantspecific Design Control Document 

(DCD) Tier 2 information). Pursuant to 

the provisions of 10 CFR 52.63(b)(1), 

also requested is an exemption from 

elements of the design as certified in the 

10 CFR part 52, appendix D, design 

certification rule for the plant-specific 

DCD departures. 

The requested amendment proposes 

changes to COL Appendix C (and plantspecific Tier 1) to reflect a new design 

of containment sump level sensors that 

affects the acceptance criterion for the 

detected containment sump level 

change test and the associated minimum 

detectable unidentified leakage rate in 

plant-specific DCD Tier 2 information. 

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 is to the containment 

sump water level instrumentation and its 

expected [reactor coolant system (RCS)] 

leakage detection capability. The affected 

equipment is not safety-related, but the 

containment sump water level sensors are 

seismically qualified. The change in 

containment sump level monitoring 

instruments has no adverse effect on the 

ability to detect a 0.5 [gallons per minute 

(gpm)] leak in containment, and therefore, 

has no adverse effect on design criteria for 

leak-before-break. The change does not affect 

the operation of any systems or equipment 

that initiate an analyzed accident or alter any 

structures, systems, and components (SSC) 

accident initiator or initiating sequence of 

events. 

Because the containment sump water level 

monitoring channels are still capable of 

detecting a 0.5 gpm leak in containment, the 

change to the SSC has no effect on plant 

operations. There is no change to plant 

systems or the response of systems to 

postulated accident conditions. There is no 

change to the predicted radioactive releases 

due to normal operation or postulated 

accident conditions. 

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 change does not affect the 

operation of any systems or equipment that 

may initiate a new or different kind of 

accident, or alter any SSC such that a new 

accident initiator or initiating sequence of 

events is created. The proposed change to the 

containment sump water level 

instrumentation and its expected RCS 

leakage detection capability has no adverse 

effect on the ability to detect a 0.5 gpm leak 

in containment. The containment sump level 

instrumentation functions are unchanged and 

leak-before-break design criteria are not 

adversely affected. 

Loss of coolant accidents for a spectrum of 

pipe sizes and locations are already 

postulated in UFSAR Chapter 15, Section 

15.6. Breaks in the main steam lines inside 

containment are also analyzed in UFSAR 

Chapter 15, Section 15.1. Unidentified 

leakage detection and operator action in 

response to unidentified leakage are not 

postulated for any of the design basis 

accident analyses described in UFSAR 

Chapter 15. 

Therefore, the requested amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The described change to the containment 

sump water level instrumentation and its 

expected RCS leakage detection capability is 

proposed to verify that the ability to detect 

a 0.5 gpm leak in containment is maintained. 

The proposed change does not alter any 

safety-related equipment, applicable design 

codes, code compliance, design function, or 

safety analysis. By ensuring that the chosen 

equipment can detect a 0.5 gpm leak in 

containment with the described accuracy, 

guidance in Regulatory Guide 1.45, Revision 

0, as committed to in the UFSAR, and 

requirements in the Technical Specifications 

are met which ensures that leak-before-break 

design criteria are not adversely affected. 

Consequently, no safety analysis or design 

basis acceptance limit/criterion is challenged 

or exceeded by the proposed change, thus the 

margin of safety is not reduced. 

Therefore, the proposed amendment 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 hazard consideration 

Attorney for licensee: M. Stanford 

Blanton, Balch & Bingham LLP, 1710 

Sixth Avenue, North, Birmingham, AL 

35203–2015. 

NRC Branch Chief: Jennifer L. DixonHerrity. 

Southern Nuclear Operating Company, 

Docket Nos. 52–025 and 52–026, Vogtle 

Electric Generating Plant, Unit Nos. 3 

and 4, Burke County, Georgia 

Date of amendment request: April 27, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18117A464. 

Description of amendment request: 

The requested amendment proposes to 

depart from Tier 2 information in the 

Updated Final Safety Analysis Report 

(UFSAR) (which includes the plantspecific Design Control Document Tier 

2 information) and involves related 

changes to plant-specific Tier 1 

information, with corresponding 

changes to the associated combined 

license (COL) Appendix C information. 

Specifically, the amendment, if 

approved, would revise the Tier 2 

information in the UFSAR and related 

changes to Tier 1 and the associated 

COL Appendix C to remove the fire 

protection system non-safety related 

containment cable spray and install 

passive fire stops and radiant energy 

shields. The changes to Tier 1 require an 

exemption, which is included in the 

license amendment request. 

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 do not affect the 

operation or reliability of any system, 

structure or component (SSC) required to 

maintain a normal power operating condition 

or to mitigate anticipated transients without 

safety-related systems. Testing has 

demonstrated that the passive fire stops 

prevent propagation of fires along the length 

of cable trays and prevent the propagation of 

cable tray fires to adjacent fire zones. The 

proposed changes do not affect the operation 

of equipment whose failure could initiate an 

accident previously analyzed. The existence 

or failure of passive fire stops in fire zone 

1100 AF 11300B does not affect normal 

equipment operation. 


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The proposed changes do not adversely 

affect the reliability or function of an SSC 

relied upon to mitigate an accident 

previously analyzed. The existence or failure 

of passive fire stops in fire zone 1100 AF 

11300B will not adversely affect passive core 

cooling system (PXS) performance during 

containment recirculation because the 

passive fire stops are located outside of the 

zone of influence (ZOI) of postulated high 

energy line breaks, and the passive fire stops’ 

material-of-construction complies with incontainment refueling water storage tank 

(IRWST) and containment recirculation 

screens design criteria for debris generation 

and transport. 

The existing active open nozzle cable tray 

suppression system is not fully automatic, is 

nonsafety-related, and is not credited in the 

probabilistic risk assessment (PRA). 

Therefore, replacing the active open nozzle 

cable tray suppression system with passive 

fire stops does not have an impact on PRA 

calculations and results. 

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 do not affect the 

operation of systems or equipment that could 

initiate a new or different kind of accident, 

or alter any SSC such that a new accident 

initiator or initiating sequence of events is 

created. The use of passive fire stops is 

recognized by Regulatory Guide 1.189. The 

passive fire stops in nonsafety-related open 

cable trays are more reliable than active 

systems such as the current open nozzle 

cable tray suppression system because they 

require no mechanical or human action to 

perform their protective function. When 

protection is required, there is no delay for 

operator or mechanical response. Testing has 

demonstrated that the passive fire stops 

prevent propagation of fires along the length 

of cable trays and prevent the propagation of 

cable tray fires to adjacent fire zones. 

The existence or failure of passive fire 

stops in fire zone 1100 AF 11300B will not 

adversely affect passive core cooling system 

(PXS) performance during containment 

recirculation because the passive fire stops 

are located outside of the zone of influence 

(ZOI) of postulated high energy line breaks, 

and their material-of-construction complies 

with in-containment refueling water storage 

tank (IRWST) and containment recirculation 

screens design criteria for debris generation 

and transport. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not affect 

existing safety margins. The current open 

nozzle cable tray suppression system is 

nonsafety-related. The use of passive fire 

stops is recognized by Regulatory Guide 

1.189. The passive fire stops in nonsafetyrelated open cable trays are more reliable 

than active systems such as the current open 

nozzle cable tray suppression system because 

they require no mechanical or human action 

to perform their protective function. When 

protection is required, there is no delay for 

operator or mechanical response. Testing has 

demonstrated that the passive fire stops 

prevent propagation of fires along the length 

of cable trays and prevent the propagation of 

cable tray fires to adjacent fire zones. 

Therefore, the proposed amendment 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 hazard consideration. 

Attorney for licensee: M. Stanford 

Blanton, Balch & Bingham LLP, 1710 

Sixth Avenue, North, Birmingham, AL 

35203–2015. 

NRC Branch Chief: Jennifer L. DixonHerrity. 

Southern Nuclear Operating Company, 

Docket Nos. 52–025 and 52–026, Vogtle 

Electric Generating Plant, Unit Nos. 3 

and 4, Burke County, Georgia 

Date of amendment request: April 27, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18117A464. 

Description of amendment request: 

The requested amendment proposes to 

depart from Tier 2 information in the 

Updated Final Safety Analysis Report 

(UFSAR) (which includes the plantspecific Design Control Document Tier 

2 information) and involves related 

changes to plant-specific Tier 1 

information, with corresponding 

changes to the associated combined 

license (COL) Appendix C information. 

Specifically, the amendment, if 

approved, would revise the Tier 2 

information in the UFSAR and related 

changes to Tier 1 and the associated 

COL Appendix C to remove the fire 

protection system non-safety related 

containment cable spray and install 

passive fire stops and radiant energy 

shields. The changes to Tier 1 require an 

exemption, which is included in the 

license amendment request. 

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 do not affect the 

operation of any systems or equipment that 

initiate an analyzed accident or alter any 

structures, systems, and components (SSC) 

accident initiator or initiating sequence of 

events. 

The proposed changes do not affect the 

physical design and operation of the Passive 

Residual Heat Removal Heat Exchanger 

(PRHR HX) or In-containment Refueling 

Water Storage Tank (IRWST) as described in 

the Updated Final Safety Analysis Report 

(UFSAR). The proposed changes do not affect 

the probability of inadvertent operation or 

failure. Therefore, the probabilities of the 

accidents previously evaluated in the UFSAR 

are not affected. 

The proposed changes do not affect the 

ability of the PRHR HX and IRWST to 

perform their design functions. The designs 

of the PRHR HX and IRWST continue to meet 

the same regulatory acceptance criteria, 

codes, and standards as required by the 

UFSAR. In addition, the proposed changes 

maintain the capabilities of the PRHR HX 

and IRWST to mitigate the consequences of 

an accident and to meet the applicable 

regulatory acceptance criteria. 

The proposed changes do not affect the 

prevention and mitigation of other abnormal 

events (e.g. anticipated operational 

occurrences, earthquakes, floods and turbine 

missiles), or their safety or design analyses. 

Therefore, the consequences of the accidents 

evaluated in the UFSAR are not affected. 

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 do not affect the 

operation of any systems or equipment that 

may initiate a new or different kind of 

accident, or alter any SSC such that a new 

accident initiator or initiating sequence of 

events is created. 

The proposed changes do not affect any 

other SSC design functions or methods of 

operation in a manner that results in a new 

failure mode, malfunction, or sequence of 

events that affect safety-related or nonsafety 

related equipment. Therefore, this activity 

does not allow for a new fission product 

release path, result in a new fission product 

barrier failure mode, or create a new 

sequence of events that result in significant 

fuel cladding failures. 

Therefore, the requested amendment does 

not create the possibility of a new or different 

type 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 changes maintain existing 

safety margins. The proposed changes verify 

and maintain the capabilities of the PRHR 

HX and IRWST to perform their design 

functions. Therefore, the proposed changes 


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satisfy the same design functions in 

accordance with the same codes and 

standards as stated in the UFSAR. These 

changes do not affect any design code, 

function, design analysis, safety analysis 

input or result, or design/safety margin. 

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 requested amendment 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: M. Stanford 

Blanton, Balch & Bingham LLP, 1710 

Sixth Avenue North, Birmingham, AL 

35203–2015. 

NRC Branch Chief: Jennifer L. DixonHerrity. 

Virginia Electric and Power Company, 

Docket Nos. 50–280 and 50–281, Surry 

Power Station, Unit Nos. 1 and 2 

(Surry), Surry County, Virginia 

Date of amendment request: March 2, 

2018. A publicly-available version is in 

ADAMS under Accession No. 

ML18075A021. 

Description of amendment request: 

The amendments would revise the 

Technical Specifications (TSs) 

consistent with Revision 0 to the 

Technical Specifications Task Force 

(TSTF) Standard Technical 

Specification Change Document TSTF– 

490, ‘‘Deletion of E Bar Definition and 

Revision to RCS Specific Activity Tech 

Spec.’’ The proposed amendments 

would adopt TSTF–490 and make the 

following associated changes: (1) 

Adoption of a TS change to replace the 

current limits on primary coolant gross 

specific activity with limits on primary 

coolant noble gas activity, and (2) an 

update of the Alternative Source Term 

(AST) analyses for Surry. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

Criterion 1. The Proposed Changes Do Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

Reactor coolant specific activity is not an 

initiator for any accident previously 

evaluated, and the allowed time period when 

primary coolant gross activity is not within 

limits is not an initiator for any accident 

previously evaluated. In addition, the current 

variable limit on primary coolant iodine 

concentration is not an initiator to any 

accident previously evaluated. Updating the 

Alternative Source Term analyses does not 

require any changes to any plant structures, 

systems, or components (SSCs) and therefore 

does not affect any accident initiators. As a 

result, the proposed changes do not 

significantly increase the probability of an 

accident. The proposed TS change will limit 

primary coolant noble gases to 

concentrations consistent with the accident 

analyses, and the proposed completion time 

when the limit may be exceeded has no 

impact on the consequences of any design 

basis accident since the consequences of an 

accident during this time period is the same 

as the consequences of an accident during 

the existing time periods. The revised 

assessments of the radiological consequences 

due to design basis accidents listed in the 

Surry Updated Final Safety Analysis Report, 

using the updated AST methodology and 

proposed assumptions and inputs, conclude 

that the Exclusion Area Boundary (EAB), 

Low Population Zone (LPZ), and Control 

Room doses are within the limits of 10 CFR 

50.67 and within the limits of Regulatory 

Guide (RG) 1.183. As a result, the 

consequences of any accident previously 

evaluated are not significantly increased. 

Criterion 2. The Proposed Changes Do Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated 

The proposed TS change in specific 

activity limits and the updated AST dose 

consequences analyses do not alter any 

physical part of the plant, (i.e., no new or 

different type of equipment will be installed,) 

nor do they affect any plant operating 

parameter or create new accident precursors. 

Therefore, the proposed changes do not 

create the potential for a new or different 

kind of accident from any previously 

calculated. 

Criterion 3. The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety. 

The proposed TS change in specific 

activity limits is consistent with the 

assumptions in -the safety analyses and will 

ensure the monitored values protect the 

initial assumptions in the safety analyses. 

The proposed changes for radiological events 

related to the computer code used to 

calculate dose, revised X/Qs for control room 

and offsite receptors (including the computer 

code and method used to determine control 

room X/Qs for SG releases), the computer 

code used to determine core inventory, the 

change in FHA [Fuel Handling Accident] gap 

fraction methodology, and removing the LRA 

[Locked Rotor Accident] from the 

radiological design basis have been analyzed 

and result in acceptable consequences, 

meeting the criteria as specified in 10 CFR 

50.67 and RG 1.183. The proposed changes 

will not result in plant operation in a 

configuration outside the analyses or design 

basis and do not adversely affect systems that 

are required to respond for safe shutdown of 

the plant and to maintain the plant in a safe 

operating condition. Therefore, the 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: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., 120 Tredegar 

St., RS–2, Richmond, VA 23219. 

NRC Branch Chief: Michael T. 

Markley. 

III. Notice of Issuance of Amendments 

to Facility Operating Licenses and 

Combined Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

of 1954, as amended (the Act), and the 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR chapter I, which are set forth in 

the license amendment. 

A notice of consideration of issuance 

of amendment to facility operating 

license or combined license, as 

applicable, proposed no significant 

hazards consideration determination, 

and opportunity for a hearing in 

connection with these actions, was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

amendment, (2) the amendment, and (3) 

the Commission’s related letter, Safety 

Evaluation, and/or Environmental 

Assessment as indicated. All of these 

items can be accessed as described in 

the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 


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DTE Electric Company, Docket No. 50– 

341, Fermi 2, Monroe County, Michigan 

Date of amendment request: August 

14, 2017. 

Brief description of amendment: The 

amendment modified Fermi 2 Technical 

Specification 5.5.7, ‘‘Ventilation Filter 

Testing Program (VFTP),’’ by adopting 

the format and language of NUREG– 

1433, ‘‘Standard Technical 

Specifications for General Electric BWR/ 

4 Plants,’’ Revision 4. 

Date of issuance: May 24, 2018. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No: 208. A publiclyavailable version is in ADAMS under 

Accession No. ML18108A022; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. NPF–43: The amendment revised 

the Facility Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: September 26, 2017 (82 FR 

44851). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated May 24, 2018. 

No significant hazards consideration 

comments received: No. 

Duke Energy Progress, LLC, Docket Nos. 

50–325 and 50–324, Brunswick Steam 

Electric Plant, Unit Nos. 1 and 2, 

Brunswick County, North Carolina 

Date of amendment request: June 29, 

2017, as supplemented by letters dated 

January 4, 2018, and January 23, 2018. 

Brief description of amendments: The 

amendments adopted Technical 

Specifications Task Force (TSTF) 

Traveler TSTF–542, Revision 2, 

‘‘Reactor Pressure Vessel Water 

Inventory Control,’’ for Brunswick 

Steam Electric Plant, Units 1 and 2. The 

amendments replaced existing technical 

specification (TS) requirements 

associated with ‘‘operations with the 

potential for draining the reactor 

vessel,’’ with revised TSs providing 

alternative requirements for reactor 

pressure vessel water inventory control. 

These alternative requirements protect 

Safety Limit 2.1.1.3, which states, 

‘‘Reactor vessel water level shall be 

greater than the top of active irradiated 

fuel.’’ 

Date of issuance: April 13, 2018. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to the 2019 Unit 2 refueling 

outage. This Notice of Issuance corrects 

the effective date of License 

Amendment No. 283, originally noticed 

in the Federal Register on May 8, 2018 

(83 FR 20865). 

Amendment Nos.: 283 (Unit 1) and 

311 (Unit 2). A publicly-available 

version is in ADAMS under Accession 

No. ML18039A444; documents related 

to this amendment are listed in the 

Safety Evaluation enclosed with the 

amendments. Amendment Nos. 283 and 

311 were corrected by letter dated May 

23, 2018 (ADAMS Accession No. 

ML18137A143). 

Renewed Facility Operating License 

No. DPR–49: The amendments revised 

the Renewed Facility Operating 

Licenses and TSs. 

Date of initial notice in Federal 

Register: September 12, 2017 (82 FR 

42846). The supplemental letters dated 

January 4, 2018, and January 23, 2018, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

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 April 13, 2018. 

No significant hazards consideration 

comments received: No. 

Duke Energy Progress, LLC, Docket No. 

50–261, H. B. Robinson Steam Electric 

Plant, Unit No. 2, Darlington County, 

South Carolina 

Date of amendment request: April 3, 

2017, as supplemented by letters dated 

April 3, 2017; May 2, 2017; September 

28, 2017; and January 8, 2018. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications (TSs) to extend the 

required frequency of certain 18-month 

Surveillance Requirements to 24 months 

to accommodate a 24-month refueling 

cycle. In addition, the amendment 

revised certain programs in TS Section 

5.5, ‘‘Programs and Manuals,’’ to change 

18-month frequencies to 24 months. 

Date of issuance: May 25, 2018. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the end of the 

next refueling outage. 

Amendment No.: 258. A publiclyavailable version is in ADAMS under 

Accession No. ML18115A150; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–23: The amendment revised 

the Renewed Facility Operating License 

and TSs. 

Date of initial notice in Federal 

Register: July 5, 2017 (82 FR 31092). 

The supplemental letters dated 

September 28, 2017, and January 8, 

2018, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the NRC 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 May 25, 2018. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1 (Clinton), DeWitt 

County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station (LaSalle), Unit Nos. 1 

and 2, LaSalle County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station (Limerick), 

Unit Nos. 1 and 2, Montgomery County, 

Pennsylvania 

Exelon Generation Company, LLC, 

Docket No. 50–410, Nine Mile Point 

Nuclear Station, Unit No. 2 (Nine Mile), 

Oswego County, New York 

Date of amendment request: 

November 8, 2017. 

Brief description of amendments: The 

amendments revised the technical 

specification requirements for 

secondary containment. 

Date of issuance: May 29, 2018. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: Clinton—218; 

LaSalle, Units 1 and 2—228 and 214; 

Limerick, Units 1 and 2—229 and 192; 

and Nine Mile—169. A publiclyavailable version is in ADAMS under 

Accession No. ML18113A045. 

Documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Operating License Nos. NPF– 

62, NPF–11, NPF–18, NPF–39, NPF–85, 

and NPF–69: The amendments revised 

the Facility Operating Licenses and 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 19, 2017 (82 FR 

60227). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated May 29, 2018. 

No significant hazards consideration 

comments received: No. 


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Southern Nuclear Operating Company, 

Inc.; Georgia Power Company; 

Oglethorpe Power Corporation; 

Municipal Electric Authority of Georgia; 

and City of Dalton, Georgia, Docket Nos. 

50–321 and 50–366, Edwin I. Hatch 

Nuclear Plant, Unit Nos. 1 and 2, 

Appling County, Georgia 

Date of amendment request: April 20, 

2017, as supplemented by letters dated 

September 14, 2017; February 19, 2018; 

and May 1, 2018. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications by replacing the existing 

requirements related to ‘‘operations 

with a potential for draining the reactor 

vessel’’ with new requirements on 

Reactor Pressure Vessel Water Inventory 

Control to protect Safety Limit 2.1.1.3, 

which requires reactor vessel water 

level to be greater than the top of active 

irradiated fuel. 

Date of issuance: May 31, 2018. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to the commencement of the Unit 

No. 2 refueling outage (U2R25) in 

February 2019. 

Amendment Nos.: Unit 1—290, Unit 

2—235. A publicly-available version is 

in ADAMS under Accession No. 

ML18123A368; documents related to 

these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Renewed Facility Operating License 

Nos. DPR–57 and NPF–5: The 

amendments revised the Renewed 

Facility Operating Licenses and 

Technical Specifications. 

Date of initial notice in Federal 

Register: August 29, 2017 (82 FR 

41071). The supplemental letters dated 

September 14, 2017; February 19, 2018; 

and May 1, 2018, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC 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 May 31, 2018. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority (TVA) 

Docket Nos. 50–259, 50–260, 50–296, 

and 72–052, Browns Ferry Nuclear 

Plant, Unit Nos. 1, 2, and 3, Limestone 

County, Alabama 

TVA Docket Nos. 50–327, 50–328, and 

72–034, Sequoyah Nuclear Plant, Unit 

Nos. 1 and 2, Hamilton County, 

Tennessee 

TVA Docket Nos. 50–390, 50–391, and 

72–1048, Watts Bar Nuclear Plant, Unit 

Nos. 1 and 2, Rhea County, Tennessee 

Date of amendment request: January 

4, 2017, as supplemented by letters 

dated July 7, 2017, and July 27, 2017. 

(Note: This Notice of Issuance corrects 

the amendments by adding the 

supplement dated July 27, 2017, which 

was inadvertently omitted from the 

original Federal Register notice (January 

16, 2018; 83 FR 2234). 

Brief description of amendments: The 

amendments revised TVA Emergency 

Plans for the above nuclear plants. 

Specifically, the amendments adopted 

the NRC-endorsed Radiological 

Emergency Plan Emergency Action 

Level schemes developed by the 

Nuclear Energy Institute (NEI 99–01, 

Revision 6, ‘‘Development of Emergency 

Action Levels for Non-Passive 

Reactors’’). 

Date of issuance: December 22, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 180 days from the date of its 

issuance, or July 3, 2018, whichever 

comes later. 

Amendment Nos.: Browns Ferry 

Nuclear Plant—303 (Unit 1), 327 (Unit 

2), and 287 (Unit 3); Sequoyah Nuclear 

Plant—339 (Unit 1) and 332 (Unit 2); 

and Watts Bar Nuclear Plant—118 (Unit 

1) and 18 (Unit 2). A publicly-available 

version is in ADAMS under Accession 

No. ML17289A032; documents related 

to these amendments are listed in the 

Safety Evaluations enclosed with the 

amendments. These amendments were 

corrected by letter dated May 29, 2018 

(ADAMS Accession No. ML18138A452). 

Renewed Facility Operating License 

Nos. DPR–33, DPR–52, DPR–68, DPR– 

77, and DPR–79, and Facility Operating 

License Nos, NPF–90 and NPF–96: The 

amendments revised the licenses. 

Date of initial notice in Federal 

Register: June 19, 2017 (82 FR 27891). 

The supplemental letters dated July 7, 

2017, and July 27, 2017, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 22, 

2017. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit No. 1, 

Callaway County, Missouri 

Date of amendment request: April 6, 

2017, as supplemented by letter dated 

February 5, 2018. 

Brief description of amendment: The 

amendment revised the Final Safety 

Analysis Report to clearly describe 

conformance with NRC Regulatory 

Guide 1.106, Revision 1, ‘‘Thermal 

Overload Protection for Electric Motors 

on Motor-Operated Valves.’’ 

Date of issuance: May 30, 2018. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 218. A publiclyavailable version is in ADAMS under 

Accession No. ML18124A026; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. NPF–30: The amendment revised 

the Final Safety Analysis Report. 

Date of initial notice in Federal 

Register: July 18, 2017 (82 FR 32885). 

The supplemental letter dated February 

5, 2018, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC 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 May 30, 2018. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 6th day 

of June 2018. 

For the Nuclear Regulatory Commission. 

Tara Inverso, 

Acting Deputy Director, Division of Operating 

Reactor Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2018–12506 Filed 6–18–18; 8:45 am] 

BILLING CODE 7590–01–P 


daltland on DSKBBV9HB2PROD with NOTICES