Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 161 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g., 

braille, large print), please notify 

Kimberly Meyer, NRC Disability 

Program Manager, at 301–287–0739, by 

videophone at 240–428–3217, or by 

email at Kimberly.Meyer-Chambers@

nrc.gov. Determinations on requests for 

reasonable accommodation will be 

made on a case-by-case basis. 

* * * * * 

Members of the public may request to 

receive this information electronically. 

If you would like to be added to the 

distribution, please contact the Nuclear 

Regulatory Commission, Office of the 

Secretary, Washington, DC 20555 (301– 

415–1969), or email Patricia.Jimenez@

nrc.gov or Jennifer.BorgesRoman@

nrc.gov. 

December 28, 2017. 

Denise L. McGovern, 

Policy Coordinator, Office of the Secretary. 

[FR Doc. 2017–28396 Filed 12–28–17; 4:15 pm] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2017–0238] 

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 December 5, 

2017, to December 18, 2017. The last 

biweekly notice was published on 

December 19, 2017. 

DATES: Comments must be filed by 

February 1, 2018. A request for a 

hearing must be filed by March 5, 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–2017–0238. Address 

questions about NRC dockets to Carol 

Gallagher; telephone: 301–415–3463; 

email: Carol.Gallagher@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: OWFN–2– 

A13, 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: 

Lynn Ronewicz, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington DC 

20555–0001; telephone: 301–415–1927, 

email: Lynn.Ronewicz@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2017– 

0238, 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–2017–0238. 

• 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–2017– 

0238, 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 

§ 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 


daltland on DSKBBV9HB2PROD with NOTICES
162 Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period if circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. If 

the Commission takes action prior to the 

expiration of either the comment period 

or the notice period, it will publish in 

the Federal Register a notice of 

issuance. If the Commission makes a 

final no significant hazards 

consideration determination, any 

hearing will take place after issuance. 

The Commission expects that the need 

to take this action will occur very 

infrequently. 

A. Opportunity To Request a Hearing 

and Petition for Leave To Intervene 

Within 60 days after the date of 

publication of this notice, any persons 

(petitioner) whose interest may be 

affected by this action may file a request 

for a hearing and petition for leave to 

intervene (petition) with respect to the 

action. Petitions shall be filed in 

accordance with the Commission’s 

‘‘Agency Rules of Practice and 

Procedure’’ in 10 CFR part 2. Interested 

persons should consult a current copy 

of 10 CFR 2.309. The NRC’s regulations 

are accessible electronically from the 

NRC Library on the NRC’s 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 federally 

recognized Indian Tribe, or agency 

thereof does not need to address the 

standing requirements in 10 CFR 

2.309(d) if the facility is located within 

its boundaries. Alternatively, a State, 

local governmental body, 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 


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Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 163 

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 

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. 

Dominion Nuclear Connecticut, Inc. 

(DNC), Docket No. 50–336, Millstone 

Power Station, Unit No. 2, New London 

County, Connecticut 

Date of amendment request: October 

4, 2017. A publicly-available version is 

in ADAMS under Accession No. 

ML17284A179. 

Description of amendment request: 

The amendment would revise the 

Millstone Power Station, Unit No. 2 

(MPS2) Technical Specification (TS) 

6.19, ‘‘Containment Leakage Rate 

Testing Program,’’ by replacing the 

reference to Regulatory Guide (RG) 

1.163, ‘‘Performance-Based Containment 

Leak-Test Program,’’ with a reference to 

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,’’ and the 

limitations and conditions specified in 

NEI 94–01, Revision 2–A, as the 


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implementing documents used to 

develop the MPS2 performance-based 

leakage testing program in accordance 

with 10 CFR, Appendix J, Option B, 

‘‘Primary Reactor Containment Leakage 

Testing for Water-Cooled Power 

Reactors.’’ The amendment would allow 

DNC to extend the Type A primary 

containment integrated leak rate test 

interval (ILRT) for MPS2 to 15 years and 

the Type C local leak rate test interval 

to 75 months, and incorporates the 

regulatory positions stated in RG 1.163. 

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 amendment involves 

changes to the MPS2 Containment Leakage 

Rate Testing Program. The proposed 

amendment does not involve a physical 

change to the plant or a change in the manner 

in which the plant is operated or controlled. 

The primary containment function is 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 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 any accident 

precursors or initiators. 

Therefore, the probability of occurrence of 

an accident previously evaluated is not 

significantly increased by the proposed 

amendment. 

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

3–A, and the limitations and conditions 

specified in NEI 94–01, Rev. 2–A, for 

development of the MPS2 performance-based 

leakage testing program. 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 potential consequences of 

extending the ILRT interval to 15 years have 

been evaluated by analyzing the resulting 

changes in risk. The increase in risk in terms 

of person-rem [roentgen equivalent man] per 

year within 50 miles resulting from design 

basis accidents was estimated to be 

acceptably small and determined to be 

within the guidelines published in RG 1.174. 

Additionally, the proposed change maintains 

defense-in-depth by preserving a reasonable 

balance among prevention of core damage, 

prevention of containment failure, and 

consequence mitigation. DNC has determined 

that the increase in Conditional Containment 

Failure Probability due to the proposed 

change is very small. 

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

3–A, and the limitations and conditions 

specified in NEI 94–01, Rev. 2–A, for 

development of the MPS2 performance-based 

leakage testing program, and establishes a 15- 

year interval for Type A testing and an 

interval not to exceed 75 months for Type C 

testing. The containment and the testing 

requirements to periodically demonstrate the 

integrity of the containment exist to ensure 

the plant’s ability to mitigate the 

consequences of an accident; do not involve 

any accident precursors or initiators. The 

proposed change does not involve a physical 

change to the plant (i.e., no new or different 

type of equipment will be installed) or a 

change to the manner in which the plant is 

operated or controlled. 

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 the margin of 

safety? 

Response: No. 

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

3–A, and the limitations and conditions 

specified in NEI 94–01, Rev. 2–A, for the 

development of the MPS2 performance-based 

leakage testing program, and establishes a 15- 

year interval for Type A testing and an 

interval not to exceed 75 months for Type C 

testing. This amendment does not alter the 

manner in which safety limits, limiting safety 

system setpoints, or limiting conditions for 

operation are determined. The specific 

requirements and conditions of the 

Containment Leakage Rate Testing Program, 

as defined in the TS, ensure that the degree 

of primary containment structural integrity 

and leak-tightness that is considered in the 

plant’s safety analysis is maintained. The 

overall containment leakage rate limit 

specified by the TS is maintained, and the 

Type A, Type B, and Type C containment 

leakage tests will be performed at the 

frequencies established in accordance with 

the NRC-accepted guidelines of NEI 94–01, 

Revision 3–A, and the limitations and 

conditions specified in NEI 94–01, Rev. 2–A. 

Containment inspections performed in 

accordance with other plant programs serve 

to provide a high degree of assurance that the 

containment will not degrade in a manner 

that is not detectable by an ILRT. A risk 

assessment using the current MPS2 PRA 

[probabilistic risk assessment] model 

concluded that extending the ILRT test 

interval from 10 years to 15 years results in 

a small change to the MPS2 risk profile. 

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: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Energy, Inc., 120 Tredegar Street, RS–2, 

Richmond, VA 23219. 

NRC Branch Chief: James G. Danna. 

DTE Electric Company, Docket No. 50– 

341, Fermi 2, Monroe County, Michigan 

Date of amendment request: August 

24, 2017. A publicly-available version is 

in ADAMS under Accession No. 

ML17237A176. 

Description of amendment request: 

The proposed amendment revises 

Technical Specification (TS) 3.3.1.1, 

‘‘Reactor Protection System (RPS) 

Instrumentation,’’ to eliminate the main 

steam line radiation monitor (MSLRM) 

functions for initiating (1) a reactor 

protection system automatic reactor trip 

and (2) the associated (Group 1) primary 

containment isolation system (PCIS) 

isolations, which include automatic 

closure of the main steam isolation 

valves (MSIV) and main steam line 

(MSL) drain valves. The proposed 

changes also remove requirements for 

Group 1 PCIS isolation from TS 3.3.6.1, 

‘‘Primary Containment Isolation 

Instrumentation.’’ This submittal also 

proposes the addition of two new TS 

Limiting Conditions for Operation, 

3.3.7.2 and 3.3.7.3, for the mechanical 

vacuum pump and gland seal exhauster 

trip instrumentation that will be 

required to actuate in response to high 

MSL radiation. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes eliminate the 

MSLRM trip and isolation functions from 

initiating an automatic reactor scram and 

automatic closure of the MSIVs. The 

justification for eliminating the MSLRM trip 

and MSIV isolation functions is based on the 

NRC-approved evaluation provided in GE 

LTR [General Electric Licensing Topical 

Report] NEDO–31400A, ‘‘Safety Evaluation 

for Eliminating the Boiling Water Reactor 

Main Steam Line Isolation Valve Closure 

Function and Scram Function of the Main 

Steam Line Radiation Monitor,’’ dated 

October 1992. 


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Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 165 

The MSLRM high radiation RPS scram 

function has never been credited to shut 

down the reactor in response to a postulated 

CRDA [control rod drop accident]; instead, 

the neutron monitoring system will continue 

to be the credited means to shut down the 

reactor in response to the high flux condition 

that results from the reactivity inserted by the 

CRDA. 

The consequences of an accident 

previously evaluated, have been re-evaluated 

consistent with RG [Regulatory Guide] 1.183 

Rev. 0 AST [alternate source term] (10 CFR 

50.67) for the applicable DBA [design basis 

accident] (i.e., the CRDA) as stipulated in 

NEDO–31400A. The supporting dose 

analyses demonstrate that, with continued 

credit for the automatic trip/isolation of the 

MVPs [mechanical vacuum pump] as well as 

a new proposed automatic trip of the GSEs 

[gland seal exhauster], the consequences of 

the accident are within the regulatory 

acceptance criteria recommended in RG 

1.183 Rev. 0 for compliance with 10 CFR 

50.67. As a result, the consequences of any 

accident previously evaluated are not 

significantly increased. 

The proposed modification of the trip logic 

for the MVPs to utilize the safety-related 

MSLRM signals is an improvement over the 

current licensed configuration of the MVP 

trip, which utilizes the nonsafety-related 

offgas 2-minute delay pipe radiation monitor 

‘‘High-High’’ radiation signal. Reliance on the 

safety-related MSLRM signal is consistent 

with similar approved license amendments 

and, in addition to improving the quality and 

reliability of the sensing circuit, ensures the 

signal is generated at the time of earliest 

possible detection and therefore improves the 

effectiveness of the actuation. The trip 

setpoint utilized corresponds to the same 

value previously assigned for initiating MSIV 

isolation in response to the design basis 

CRDA. The offgas 2-minute delay pipe 

radiation monitor alarm function is being 

retained, with a more conservative setpoint, 

to continue to provide indication of 

increased radiation. 

Similar to the MVPs, the proposed new trip 

of the nonsafety-related GSEs is also 

necessary to ensure calculated radiological 

consequences remain within the regulatory 

acceptance limits. Reliance on the safetyrelated MSLRM signal is consistent with 

BWR [boiling water reactor] design for 

reliable tripping of the nonsafety-related 

MVPs and ensures the signal is reliably 

generated at the time of earliest possible 

detection and maximizes the effectiveness of 

the actuation. 

The proposed changes also include the 

elimination of the MSLRM isolation function 

from automatically closing the MSL drain 

valves. The contents of the MSL drain lines 

are conveyed to the main condenser. The 

evaluation of the condenser release path 

assumes that 100% of CRDA activity released 

is transported to the main condenser in 1 

second, and therefore, the transportation of 

the post-CRDA activity from the reactor 

coolant to the main condenser either via 

MSLs or MSL drain lines is inconsequential 

and is supported by the dose analyses 

performed in support of this submittal. 

Neither the MSLRMs nor the MVPs are 

postulated initiators of any accident 

previously evaluated. None of the proposed 

changes alter the probability of the 

occurrence of the CRDA initiating event. 

The loss of the GSEs is a malfunction of 

equipment considered in UFSAR [updated 

final safety analysis report] Section 15.12 

‘‘Malfunction of Turbine Gland Sealing 

System.’’ In the event that the operating 

blower malfunctions, the backup blower will 

automatically assume the gas removal 

requirements. Assuming loss of both blowers, 

vacuum will be lost in the gland steam 

condenser. No cladding perforations result 

from a malfunction of the turbine gland 

sealing system. The pressure in the gland 

steam exhaust header will increase to greater 

than atmospheric, allowing sealing steam to 

escape into the turbine building. If exhauster 

vacuum falls below a specified value, caused 

for example by loss of alternating current 

(AC) power, a vacuum switch initiates the 

closing of the live steam supply to the gland 

steam header. Above 50% to 60% reactor 

power, the turbine is self-sealing; hence, the 

packing lines would remain pressurized 

under normal operating conditions. 

The logic associated with the new trip of 

the GSEs will be designed to preserve the 

existing ability of the backup exhauster to 

automatically respond to a loss of the 

operating exhauster, in the absence of a valid 

high MSL radiation trip signal. Similar to the 

design of the RPS trip logic that is proposed 

to be eliminated, the GSE trip logic will be 

configured such that no single failure of a 

MSLRM can generate a GSE trip signal. As 

specified in the ‘‘Applicability’’ section for 

the new proposed LCO [limiting condition 

for operation] 3.3.7.3, the trip logic will be 

automatically bypassed when reactor power 

is above 10% RTP [rated thermal power] 

when the consequences postulated in 

association with a CRDA are not credible. On 

the basis of the configuration of the GSE trip 

logic, the quality of the initiating trip logic 

signal, and the short duration of normal 

operation for which the GSE trip logic will 

be active, the probability of a malfunction of 

equipment leading to the loss of the turbine 

gland sealing system is not significantly 

increased. 

The proposed changes do not increase 

system or component pressures, 

temperatures, or flowrates for systems 

designed to prevent accidents or mitigate the 

consequences of an accident. Since these 

conditions do not change, the probability of 

a process-induced failure or malfunction of a 

SSC [system, structure, or component] is not 

increased. 

The addition of MVP and GSE SRs 

[surveillance requirements] and LCOs to the 

TS enhances the reliability of these design 

functions by establishing administrative 

requirements for periodic verification of their 

operability. 

The reliance on a lower assigned MSL high 

radiation alarm setpoint of 1.5 times the full 

power N–16 background will direct the 

control room operators to diagnose and act to 

mitigate conditions associated with fuel 

damage and release sooner than the current 

alarm condition which will reduce the 

potential consequences of a postulated 

release due to a CRDA. 

On the basis of the above considerations, 

the proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not increase 

system or component pressures, 

temperatures, or flowrates. Since these 

conditions do not change, the likelihood of 

a process-induced failure or malfunction of a 

SSC not previously considered is not 

increased. 

The reliance on the MVP trip to ensure 

acceptable dose consequences following a 

postulated CRDA is consistent with the 

original plant design and licensing bases. The 

re-assignment of the initiating input for the 

MVP trip logic to the MSLRM improves the 

quality and reliability of the credited trip 

initiating logic by relying on safety-related, 

redundant components. The quality of the 

nonsafety-related trip circuit itself is 

unchanged. 

The reliance on the proposed trip of the 

GSEs is a function that is credited to ensure 

acceptable dose consequences following a 

postulated CRDA. The use of the safetyrelated redundant MSLRM signals and 

nonsafety-related trip circuit provides the 

same level of quality and reliability of the 

initiating trip logic and trip circuitry credited 

to trip the MVPs. These requirements provide 

the reliability necessary to ensure the 

assumptions of the analyzed CRDA remain 

valid. 

Both the safety-related trip logic and the 

nonsafety-related trip circuits associated with 

the MVP and GSE trips will be designed to 

include qualified electrical isolation 

necessary to ensure the nonsafety-related trip 

circuitry cannot induce failures of or affect 

the reliability of the safety-related trip logic. 

The new GSE trip will be designed to 

preserve the existing function for auto-start of 

the standby exhauster in the event that the 

plant experiences a loss of the operating 

exhauster, in the absence of a valid high MSL 

radiation trip signal. An installed automatic 

bypass of the GSE trip is actuated once steam 

flow and feedwater flow correspond to the 

same Low Power Setpoint used to disable the 

rod block function of the Rod Worth 

Minimizer during plant startup. This bypass 

will minimize the potential for the plant to 

experience a loss of both GSEs and potential 

ensuing turbine trip due to a failure of the 

new trip circuit. The status of the GSE trip 

bypass will be available to the control room 

operators and be required to be verified as a 

part of the plant general operating procedures 

for startup/shutdown. 

Adding requirements for the MVP and GSE 

trip instrumentation in the TS will ensure 

that appropriate measures and requirements 

are in place such that any release of 

radioactive material released from a gross 

fuel failure will be contained in the main 

condenser and processed through the offgas 

system in the manner credited in the plant 

analysis of the CRDA. 

The MSLRM trip and isolation functions 

being eliminated as described above are only 

applicable to the CRDA and no other event 


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in the safety analysis. The proposed changes 

are consistent with the revised safety analysis 

assumptions for a CRDA as described in this 

license amendment request. 

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 changes eliminating the 

MSLRM trip and isolation functions from 

initiating an automatic reactor scram and 

automatic closure of the MSIVs are justified 

based on the NRC-approved LTR NEDO– 

31400A and supporting dose analysis. The 

supporting dose analysis also supports the 

elimination of the MSL drain isolation 

function of the MSLRMs on the basis that 

with the valves open the source term 

associated with the analyzed release is 

directed to the main condenser the same as 

it would be via the MSLs themselves. 

The methods of analysis and assumptions 

used to evaluate the consequences of the 

applicable impacted safety analysis (i.e. the 

CRDA) are consistent with the conservative 

regulatory requirements and guidance 

identified in Section 5.1 above [this is a 

reference to ‘‘Applicable Regulatory 

Requirements/Criteria’’ in DTE August 24, 

2017, license amendment request] and 

establish estimates of the EAB [exclusion 

area boundary], LPZ [low population zone], 

and MCR [main control room] doses that 

comply with these criteria. Hence, there is 

reasonable assurance that Fermi 2, modified 

as proposed by this submittal, will continue 

to provide sufficient safety margins to 

address unanticipated events and to 

compensate for uncertainties in accident 

progression and analysis assumptions and 

parameters. 

Adding requirements for the MVP and GSE 

high MSL radiation trips in the Fermi 2 TS 

will ensure that appropriate measures and 

requirements are in place to maintain the 

operability of these functions as such that 

any release of radioactive material from a 

gross fuel failure resulting from a CRDA will 

be contained in the main condenser and 

processed through the offgas system. 

The proposed changes do not increase 

system or component pressures, 

temperatures, or flowrates for systems 

designed to prevent accidents or mitigate the 

consequences of an accident. 

The analyses performed in accordance 

with the specified NRC-approved methods 

and assumptions demonstrate that the 

removal of the trip and isolation functions as 

described will not cause a significant 

reduction in the margin of safety, as the 

resulting offsite dose consequences are being 

maintained within regulatory limits. The 

proposed changes do not exceed or alter a 

design basis or a safety limit for a parameter 

to be described or established in the UFSAR 

[updated final safety analysis report]. 

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: Jon P. 

Christinidis, DTE Energy, Expert 

Attorney—Regulatory, 688 WCB, One 

Energy Plaza, Detroit, MI 48226–1279. 

NRC Branch Chief: David J. Wrona. 

Duke Energy Progress, LLC (Duke 

Energy), Docket No. 50–400, Shearon 

Harris Nuclear Power Plant, Unit 1 

(HNP), Wake and Chatham Counties, 

North Carolina 

Duke Energy Progress, LLC, Docket No. 

50–261, H.B. Robinson Steam Electric 

Plant Unit No. 2 (RNP), Darlington 

County, South Carolina 

Date of amendment request: October 

19, 2017. A publicly-available version is 

in ADAMS under Accession No. 

ML17292A040. 

Description of amendment request: 

The proposed amendment request 

consists of five changes that would 

revise the Technical Specifications 

(TSs) to support the allowance of Duke 

Energy to self-perform core reload 

design and safety analyses. These 

changes would (1) add the NRCapproved COPERNIC Topical Report 

(TR) to the list of TRs for HNP and RNP; 

(2) relocate several TS parameters to the 

Core Operating Limits Reports for HNP 

and RNP; (3) revise the RNP TS 

Moderator Temperature Coefficient 

maximum upper limit; (4) revise the 

HNP TS definition of Shutdown Margin 

consistent with Technical Specifications 

Task Force (TSTF) Traveler TSTF–248, 

Revision 0, ‘‘Revise Shutdown Margin 

Definition for Stuck Rod Exception’’ 

(ADAMS Accession No. ML040611010); 

and (5) revise the RNP and HNP power 

distribution limits limiting condition for 

operation actions and surveillance 

requirements to allow operation of a 

reactor core designed using the DPC– 

NE–2011–P [proprietary], ‘‘Nuclear 

Design Methodology Report for Core 

Operating Limits of Westinghouse 

Reactors,’’ methodology. (A redacted 

version, designated as DPC–NE–2011, is 

publicly-available under ADAMS 

Accession No. ML16125A420.) 

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. 

COPERNIC 

The proposed change adds a topical report 

for an NRC-reviewed and approved fuel 

performance code to the list of topical reports 

in RNP and HNP Technical Specifications 

(TS), which is administrative in nature and 

has no impact on a plant configuration or 

system performance relied upon to mitigate 

the consequences of an accident. The list of 

topical reports in the TS used to develop the 

core operating limits does not impact either 

the initiation of an accident or the mitigation 

of its consequences. 

Relocate TS Parameters to the COLR 

The proposed change relocates certain 

cycle-specific core operating limits from the 

RNP and HNP TS to the Core Operating 

Limits Report (COLR). The cycle-specific 

values must be calculated using the NRC 

approved methodologies listed in the COLR 

section of the TS. Because the parameter 

limits are determined using the NRC 

methodologies, they will continue to be 

within the limit assumed in the accident 

analysis. As a result, neither the probability 

nor the consequences of any accident 

previously evaluated will be affected. 

RNP MTC TS Change 

The proposed change revises the RNP 

Technical Specification maximum upper 

Moderator Temperature Coefficient (MTC) 

limit. Revision of the MTC limit does not 

affect the performance of any equipment 

used to mitigate the consequences of an 

analyzed accident. There is no impact on the 

source term or pathways assumed in 

accidents previously assumed. No analysis 

assumptions are violated and there are no 

adverse effects on the factors that contribute 

to offsite or onsite dose as the result of an 

accident. 

HNP TSTF–248 

The proposed change revises the HNP 

Technical Specification definition of 

Shutdown Margin (SDM) consistent with 

existing NRC-approved definition. The 

proposed revision to the SDM definition will 

result in analytical flexibility for determining 

SDM. Revision of the SDM definition does 

not affect the performance of any equipment 

used to mitigate the consequences of an 

analyzed accident. There is no impact on the 

source term or pathways assumed in 

accidents previously assumed. No analysis 

assumptions are violated and there are no 

adverse effects on the factors that contribute 

to offsite or onsite dose as the result of an 

accident. 

DPC–NE–2011–P TS Changes 

The proposed change revises the RNP and 

HNP TS to allow operation of a reactor core 

designed using the DPC–NE–2011–P 

methodology. The DPC–NE–2011–P 

methodology has already been approved by 

the NRC for use at RNP and HNP. Revision 

of the TS to align with the NRC-approved 

methodology does not affect the performance 

of any equipment used to mitigate the 

consequences of an analyzed accident. There 

is no impact on the source term or pathways 

assumed in accidents previously assumed. 


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No analysis assumptions are violated and 

there are no adverse effects on the factors that 

contribute to offsite or onsite dose as the 

result of an accident. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

COPERNIC 

The proposed change adds a topical report 

for an NRC-reviewed and approved fuel 

performance code to the list of topical reports 

in HNP and RNP TS, which is administrative 

in nature and has no impact on a plant 

configuration or on system performance. The 

proposed change updates the list of NRCapproved topical reports used to develop the 

core operating limits. There is no change to 

the parameters within which the plant is 

normally operated. The possibility of a new 

or different kind of accident is not created. 

Relocate TS Parameters to the COLR 

The proposed change relocates certain 

cycle-specific core operating limits from the 

RNP and HNP TS to the COLR. No new or 

different accidents result from utilizing the 

proposed change. The changes do not involve 

a physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. In 

addition, the changes do not impose any new 

or different requirements or eliminate any 

existing requirements. The changes do not 

alter assumptions made in the safety 

analyses. The proposed changes are 

consistent with the safety analyses 

assumptions and current plant operating 

practice. 

RNP MTC TS Change 

The proposed change revises the RNP 

Technical Specification maximum upper 

MTC limit. The proposed change does not 

physically alter the plant; that is, no new or 

different type of equipment will be installed. 

Therefore the proposed change could also not 

initiate an equipment malfunction that 

would result in a new or different type of 

accident from any previously evaluated. This 

change does not create new failure modes or 

mechanisms which are not identifiable 

during testing, and no new accident 

precursors are generated. 

HNP TSTF–248 

Revising the HNP Technical Specification 

definition of SDM would not require revision 

to any SDM boron calculations. Rather, it 

would afford the analytical flexibility for 

determining SDM for a particular 

circumstance. The proposed change does not 

physically alter the plant; that is, no new or 

different type of equipment will be installed. 

Therefore the proposed change could also not 

initiate an equipment malfunction that 

would result in a new or different type of 

accident from any previously evaluated. This 

change does not create new failure modes or 

mechanisms which are not identifiable 

during testing, and no new accident 

precursors are generated. 

DPC–NE–2011–P TS Changes 

The proposed change revises the RNP and 

HNP TS to allow operation of a reactor core 

designed using the DPC–NE–2011–P 

methodology. The DPC–NE–2011–P 

methodology has already been approved by 

the NRC for use at RNP and HNP. The 

proposed change does not physically alter 

the plant, that is, no new or different type of 

equipment will be installed. Therefore the 

proposed change could also not initiate an 

equipment malfunction that would result in 

a new or different type of accident from any 

previously evaluated. Operating the reactor 

in accordance with the NRC-approved 

methodology will ensure that the core will 

operate within safe limits. This change does 

not create new failure modes or mechanisms 

which are not identifiable during testing, and 

no new accident precursors are generated. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

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. 

COPERNIC 

The proposed change adds a topical report 

for an NRC-reviewed and approved fuel 

performance code to the list of topical reports 

in HNP and RNP TS, which is administrative 

in nature and does not amend the cycle 

specific parameters presently required by the 

TS. The individual TS continue to require 

operation of the plant within the bounds of 

the limits specified in the COLR. The 

proposed change to the list of analytical 

methods referenced in the COLR does not 

impact the margin of safety. 

Relocate TS Parameters to the COLR 

The proposed change relocates certain 

cycle-specific core operating limits from the 

RNP and HNP TS to the COLR. This change 

will have no effect on the margin of safety. 

The relocated cycle-specific parameters will 

continue to be calculated using NRCapproved methodologies and will provide the 

same margin of safety as the values currently 

located in the TS. 

RNP MTC TS Change 

The proposed change revises the RNP 

Technical Specification maximum upper 

MTC limit. The MTC limit change does not 

impact the reliability of the fission product 

barriers to function. Radiological dose to 

plant operators or to the public will not be 

impacted as a result of the proposed change. 

The current Updated Final Safety Analysis 

Report (UFSAR) Chapter 15 analyses of 

record remain bounding with the proposed 

change to the maximum upper MTC limit. 

Therefore, all of the applicable acceptance 

criteria continue to be met for each of the 

analyses with the revised maximum upper 

MTC limit. 

HNP TSTF–248 

The proposed revision to the HNP 

Technical Specification definition of SDM 

does not impact the reliability of the fission 

product barriers to function. Radiological 

dose to plant operators or to the public will 

not be impacted as a result of the proposed 

change. Adequate SDM will continue to be 

ensured for all operational conditions. 

DPC–NE–2011–P TS Changes 

The proposed change revises the RNP and 

HNP TS to allow operation of a reactor core 

designed using the DPC–NE–2011–P 

methodology. As a portion of the overall 

Duke Energy methodology for cycle reload 

safety analyses, DPC–NE–2011–P has already 

been approved by the NRC for use at RNP 

and HNP. The proposed change will continue 

to ensure that applicable design and safety 

limits are satisfied such that the fission 

product barriers will continue to perform 

their design functions. Operation of the 

reactor in accordance with the DPC–NE– 

2011–P methodology will ensure the margin 

of safety is not reduced. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Kathryn B. 

Nolan, Deputy General Counsel, Duke 

Energy Corporation, 550 South Tryon 

Street, Mail Code DEC45A, Charlotte NC 

28202. 

NRC Branch Chief: Undine Shoop. 

Duke Energy Progress, LLC, Docket No. 

50–400, Shearon Harris Nuclear Power 

Plant, Unit 1, Wake and Chatham 

Counties, North Carolina 

Date of amendment request: October 

10, 2017. A publicly-available version is 

in ADAMS under Accession No. 

ML17283A159. 

Description of amendment request: 

The amendment would revise the 

Shearon Harris Nuclear Power Plant 

(HNP), Unit 1, Technical Specifications 

(TSs) to align more closely to improved 

Standard Technical Specifications for 

rod control and to the initial conditions 

in the HNP safety analyses. The 

proposed changes will delete TS action 

statement requirements that include a 

plant shutdown to address rods that are 

immovable but trippable. Revisions to 

surveillance requirements (SRs) are 

proposed to clarify actions that are not 

necessary if rods are immovable but still 

trippable. 


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168 Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 

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 license amendment 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

Response: No. 

The proposed activity will delete action 

statement 3.1.3.1.c from the HNP TS and 

amend action statement 3.1.3.1.d, SR 

4.1.1.1.1.a, and SR 4.1.1.2.a. These TS actions 

address electrical problems that prevent the 

Control Rod Drive Mechanism (CRDM) from 

moving rods. These conditions do not affect 

the safety functions of the control rods or 

shutdown margin of the unit. Rods will still 

insert into the core on an interruption of 

power to the CRDM, as occurs in a reactor 

trip. Also, rod alignment is not impacted, 

ensuring no change to reactivity. 

The proposed activity is removing actions 

from the HNP TS for conditions that do not 

impact the plant’s safety analysis. Rods will 

still insert into the core on an interruption of 

power to the CRDM, as occurs in a reactor 

trip. Also, rod alignment is not impacted, 

ensuring no change to reactivity or shutdown 

margin. Since the conditions of these TS 

actions do not impact the plant safety 

analysis, the plant shutdown directed by 

them is unnecessary. The overall probability 

or consequence of an accident will not be 

significantly increased by removing the 

unnecessary TS actions. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed license amendment 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated? 

Response: No. 

The proposed activity will delete action 

statement 3.1.3.1.c from the HNP TS and 

amend action statements 3.1.3.1.d, SR 

4.1.1.1.1.a, and SR 4.1.1.2.a. These TS actions 

address electrical problems that prevent the 

CRDM from moving rods. These conditions 

do not affect the safety functions of the 

control rods. Rods will still insert into the 

core on an interruption of power to the 

CRDM, as occurs in a reactor trip. Also, rod 

alignment is not impacted, ensuring no 

change to reactivity or shutdown margin. 

The proposed change does not involve 

installation of new equipment or 

modification of existing equipment, so that 

no new equipment failure modes are 

introduced. Also, the proposed change in TS 

does not result in a change to the way that 

the equipment or facility is operated that 

would create new accident initiators. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed license amendment 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The proposed activity will delete action 

statement 3.1.3.1.c from the HNP TS and 

amend action statement 3.1.3.1.d, SR 

4.1.1.1.1.a, and SR 4.1.1.2.a. These actions 

address electrical problems that prevent the 

CRDM from moving rods. These conditions 

do not affect the safety functions of the 

control rods. Rods will still insert into the 

core on an interruption of power to the 

CRDM, as occurs in a reactor trip. Also, rod 

alignment is not impacted, ensuring no 

change to reactivity or shutdown margin. 

The TS action statements as amended will 

continue to address the two required safety 

functions of rod control: to shut down the 

reactor in the event of a reactor trip, or to 

maintain proper alignment to ensure even 

power distribution. TS action statement 

3.1.3.1.a will remain to drive actions if 

untrippable rods are identified. TS action 

statements 3.1.3.1.b and 3.1.3.1.d will remain 

to drive actions if misaligned rods are 

identified. The proposed changes to HNP TS 

do not significantly impact either rod safety 

function, and separate TS action statements 

for both functions will remain in place. 

Further, the impacted surveillances will 

continue to be applicable to conditions 

impacting either rod safety function. 

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, Duke 

Energy Corporation, 550 South Tryon 

St., M/C DEC45A, Charlotte, NC 28202. 

NRC Branch Chief: Undine Shoop. 

Exelon Generation Company, LLC, 

Docket No. 50–244, R.E. Ginna Nuclear 

Power Plant, Wayne County, New York 

Date of amendment request: October 

31, 2017. A publicly available version is 

in ADAMS under Accession No. 

ML17304A984. 

Description of amendment request: 

The amendment would revise Technical 

Specification (TS) Surveillance 

Requirement 3.8.4.3, ‘‘DC [Direct 

Current] Sources—MODES 1, 2, 3, and 

4,’’ for the R.E. Ginna Nuclear Power 

Plant (Ginna). The proposed change 

would allow the use of a consistent 

battery testing technique in order to 

provide consistent data for trending 

battery performance. This proposed 

change is based on guidance provided 

in the Institute of Electrical and 

Electronics Engineers (IEEE) Standard 

450–2010, ‘‘IEEE Recommended 

Practice for Maintenance, Testing, and 

Replacement of Vented Lead-Acid 

Batteries for Stationary Applications,’’ 

which is endorsed by NRC Regulatory 

Guide 1.129, Revision 3, ‘‘Maintenance, 

Testing, and Replacement of Vented 

Lead-Acid Storage Batteries for Nuclear 

Power Plants.’’ 

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 does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. The proposed change will 

continue to ensure that the DC system is 

tested in a manner that will verify 

operability. Performance of the required 

system surveillances, in conjunction with the 

applicable operational and design 

requirements for the DC system, provide 

assurance that the system will be capable of 

performing the required design functions for 

accident mitigation and also that the system 

will perform in accordance with the 

functional requirements for the system as 

described in the Updated Final Safety 

Analysis Report for Ginna. This change is in 

accordance with IEEE Standard 450–2010, 

‘‘IEEE Recommended Practice for 

Maintenance, Testing, and Replacement of 

Vented Lead-Acid Batteries for Stationary 

Applications,’’ which has been endorsed by 

NRC Regulatory Guide 1.129, Revision 3, 

‘‘Maintenance, Testing, and Replacement of 

Vented Lead-Acid Storage Batteries for 

Nuclear Power Plants.’’ This endures that the 

rate of occurrence and consequences of 

analyzed accidents will not 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 does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. The 

proposed surveillance requirement change 

will continue to ensure that the DC system 

and in particular the batteries are tested in 

a manner that will verify operability. No 

physical changes to the Ginna systems, 

structures, or components are being 

implemented. There are no new or different 

accident initiators or sequences being created 

by the proposed TS change. Therefore, the 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change does not involve a 

significant reduction in the margin of safety. 


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The proposed DC system surveillance 

requirement change provides appropriate and 

applicable surveillances for the DC system. 

The proposed change to surveillance 

requirements for the DC system will continue 

to ensure system operability. 

Therefore, this change does not affect any 

margin of safety for Ginna. 

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: Tamra Domeyer, 

Associate General Counsel, Exelon 

Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: James G. Danna. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant, Units Nos. 1 and 

2, Berrien County, Michigan 

Date of amendment request: 

November 7, 2017. A publicly-available 

version is in ADAMS under Accession 

No. ML17317A472. 

Description of amendment request: 

The proposed change would allow for 

deviation from National Fire Protection 

Association (NFPA) 805 requirements to 

allow for currently installed nonplenum listed cables routed above 

suspended ceilings and to allow for the 

use of thin wall electrical metallic 

tubing (EMT) and embedded/buried 

plastic conduit. 

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 use of EMT and embedded/buried PVC 

[polyvinyl chloride] does not create ignition 

sources and does not impact fire prevention. 

The EMT and embedded PVC had been in 

use since original plant construction, are 

allowed by the National Electrical Code and 

are not expected to increase the potential for 

a fire to start. 

The prior introduction of non-listed 

communication/data cables routed above 

suspended ceilings does not create ignition 

sources and does not impact fire prevention. 

Cable installation procedures are utilized to 

prevent the future installation of new cables 

that are noncompliant. Also, the 

communication/data cables routed above 

suspended ceilings do not result in 

compromising automatic fire suppression 

functions, manual fire suppression functions, 

fire protection or systems and structures, or 

post-fire safe shutdown capability. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do allow future 

physical changes to the facility that deviate 

from NFPA 805 requirements. However, the 

proposed changes do not alter any 

assumptions made in the safety analyses, nor 

do they involve any changes to plant 

procedures for ensuring that the plant is 

operated within analyzed limits. As such, no 

new failure modes or mechanisms that could 

cause a new or different kind of accident 

from any previously evaluated 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. 

The proposed changes do not alter the 

manner in which safety limits or limiting 

safety system settings are determined. No 

changes to instrument/system actuation 

setpoints are involved. The safety analysis 

acceptance criteria are not affected by this 

change and the proposed changes will not 

permit plant operation in a configuration 

outside the design basis. 

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: Robert B. 

Haemer, Senior Nuclear Counsel, One 

Cook Place, Bridgman, MI 49106. 

NRC Branch Chief: David J. Wrona. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant (CNP), Units Nos. 

1 and 2, Berrien County, Michigan 

Date of amendment request: 

November 7, 2017. A publicly-available 

version is in ADAMS under Package 

Accession No. ML17317A454. 

Description of amendment request: 

The proposed change would revise the 

CNP Emergency Plan to relocate the 

Technical Support Center (TSC) within 

the CNP protected area. 

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 to the CNP 

emergency plan to relocate the TSC does not 

impact the physical function of plant 

structures, systems, or components (SSC) or 

the manner in which SSCs perform their 

design function. The proposed change 

neither adversely affects accident initiators or 

precursors, nor alters design assumptions. 

The proposed change does not alter or 

prevent the ability of SSCs to perform their 

intended function to mitigate the 

consequences of an initiating event within 

assumed acceptance limits. No operating 

procedures or administrative controls that 

function to prevent or mitigate accidents are 

affected by the proposed changes. 

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 impact the 

accident analysis. The proposed change does 

not involve a physical alteration of the plant 

(i.e., no new or different type of equipment 

will be installed or removed) or a change in 

the method of plant 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 to 

the location of the TSC is not an initiator of 

any accidents. 

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 amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Margin of safety is associated with 

confidence in the ability of the fission 

product barriers (i.e., fuel cladding, reactor 

coolant system pressure boundary, and 

containment structure) to limit the level of 

radiation dose to the public. The proposed 

change does not impact operation of the 

plant or its response to transients or 

accidents. The change does not affect the 

Technical Specifications or the operating 

license other than to amend the license to 

approve the change. The proposed change 

does not involve a change in the method of 

plant operation, and no accident analyses 

will be affected by the proposed changes. 

Additionally, the proposed change will not 

relax any criteria used to establish safety 

limits and will not relax any safety system 

settings. The safety analysis acceptance 

criteria are not affected by these changes. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. The proposed change does not 

adversely affect systems that respond to 

safely shut down the plant and to maintain 


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the plant in a safe shutdown condition. The 

emergency plan will continue to activate an 

emergency response commensurate with the 

extend of degradation of plant safety. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Robert B. 

Haemer, Senior Nuclear Counsel, One 

Cook Place, Bridgman, MI 49106. 

NRC Branch Chief: David J. Wrona. 

Southern Nuclear Operating Company, 

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

Electric Generating Plant (VEGP), Units 

3 and 4, Burke County, Georgia 

Date of amendment request: October 

6, 2017. A publicly-available version is 

in ADAMS under Accession No. 

ML17279B017. 

Description of amendment request: 

The requested amendment proposes 

changes to the licensing basis 

documents to change the methodology 

and acceptance criteria for the incontainment refueling water storage 

tank (IRWST) heatup preoperational test 

described in the Updated Final Safety 

Analysis Report (UFSAR) Subsection 

14.2.9.1.3, item h, and the passive 

residual heat removal (PRHR) heat 

exchanger preoperational test described 

in UFSAR Subsection 14.2.9.1.3, item g. 

These changes involve material which is 

specifically referenced in Section 2.D.(2) 

of the combined licenses for VEGP, 

Units 3 and 4. 

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. 

This activity changes the acceptance 

criteria for the IRWST heatup preoperational 

test and provides allowance to perform the 

preoperational test during both PRHR heat 

exchanger natural circulation and forced 

flow, instead of only during natural 

circulation. In addition, the acceptance 

criteria are changed for the PRHR heat 

exchanger forced flow system operability and 

preoperational tests. 

No structure, system, or component (SSC) 

or function is changed by this proposed 

activity. There is no change to the 

application of Regulatory Guide 1.68, nor is 

there a change to the design of the PRHR heat 

exchanger or the IRWST. The initial test 

program continues to confirm the heat 

transfer capability of the PRHR heat 

exchanger and that the IRWST heatup is 

consistent with the PRHR heat exchanger 

heat transfer modeling in the UFSAR Chapter 

15 safety analysis. 

The proposed amendment does not affect 

the prevention or mitigation of abnormal 

events; e.g., accidents, anticipated operation 

occurrences, earthquakes, floods, turbine 

missiles, and fires or their safety or design 

analyses. This change does not involve 

containment of radioactive isotopes or have 

any adverse effect on a fission product 

barrier. There is no impact on previously 

evaluated accidents. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

new failure mechanism or malfunction, that 

affects an SSC accident initiator, or interface 

with any SSC accident initiator or initiating 

sequence of events considered in the design 

and licensing bases. There is no adverse 

effect on radioisotope barriers or the release 

of radioactive materials. The proposed 

amendment does not adversely affect any 

accident, including the possibility of creating 

a new or different kind of accident from any 

accident previously evaluated. 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. 

This activity changes the acceptance 

criteria for the IRWST heatup preoperational 

test and gives allowance to perform the 

preoperational test during both PRHR heat 

exchanger natural circulation and forced 

flow, instead of only during natural 

circulation. In addition, the acceptance 

criteria are changed for the PRHR heat 

exchanger forced flow system operability and 

preoperational tests. 

No SSC or function is changed within this 

activity. There is no change to the 

application of Regulatory Guide 1.68, nor is 

there a change to how the PRHR heat 

exchanger or the IRWST are designed. The 

initial test program continues to confirm the 

heat transfer capability of the PRHR heat 

exchanger. The initial test program will 

confirm the IRWST heatup is consistent with 

the current PRHR heat exchanger heat 

transfer modeling in the UFSAR Chapter 15 

safety analysis. 

The proposed changes would not affect any 

safety-related design code, function, design 

analysis, safety analysis input or result, or 

existing design/safety margin. No safety 

analysis or design basis acceptance limit/ 

criterion is challenged or exceeded by the 

requested changes. 

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 DixonHerrity. 

Southern Nuclear Operating Company, 

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

Electric Generating Plant, Units 3 and 4, 

Burke County, Georgia 

Date of amendment request: 

November 16, 2017. A publiclyavailable version is in ADAMS under 

Accession No. ML17325A562. 

Description of amendment request: 

The amendments propose changes to 

Inspections, Tests, Analyses, and 

Acceptance Criteria (ITAAC) in 

Combined License (COL) Appendix C, 

with corresponding changes to the 

associated plant-specific Tier 1 

information to simplify and consolidate 

a number of ITAAC to improve 

efficiency of the ITAAC completion and 

closure process. Pursuant to the 

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

exemption from elements of the design 

as certified in the 10 CFR part 52, 

Appendix D, design certification rule is 

also requested for the plant-specific 

Design Control Document Tier 1 

material departures. 

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 non-technical change to COL 

Appendix C will consolidate ITAAC in order 

to improve and create a more efficient 

process for the ITAAC Closure Notification 

submittals. No structure, system, or 

component (SSC) design or function is 

affected. No design or safety analysis is 

affected. The proposed changes do not affect 

any accident initiating event or component 

failure, thus the probabilities of the accidents 

previously evaluated are not affected. No 

function used to mitigate a radioactive 

material release and no radioactive material 

release source term is involved, thus the 


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radiological releases in the accident analyses 

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 change to COL Appendix C 

does not affect the design or function of any 

SSC, but will consolidate ITAAC in order to 

improve efficiency of the ITAAC completion 

and closure process. The proposed changes 

would not introduce a new failure mode, 

fault or sequence of events that could result 

in a radioactive material release. 

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 change to COL Appendix C 

to consolidate ITAAC in order to improve 

efficiency of the ITAAC completion and 

closure process is considered non-technical 

and would not affect any design parameter, 

function or analysis. There would be no 

change to an existing design basis, design 

function, regulatory criterion, or analysis. No 

safety analysis or design basis acceptance 

limit/criterion is involved. 

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. 

Attorney for licensee: M. Stanford 

Blanton, Balch & Bingham LLP, 1710 

Sixth Avenue North, Birmingham, AL 

35203–2015. 

NRC Branch Chief: Jennifer DixonHerrity. 

Tennessee Valley Authority, Docket No. 

50–391, Watts Bar Nuclear Plant, Unit 2, 

Rhea County, Tennessee 

Date of amendment request: October 

11, 2017. A publicly-available version is 

in ADAMS under Accession No. 

ML17284A452. 

Description of amendment request: 

The amendment would revise Technical 

Specification (TS) 3.3.1, Table 3.3.1–1, 

‘‘Reactor Trip System (RTS) 

Instrumentation,’’ to increase the values 

for the nominal trip setpoint and the 

allowable value for Function 14.a, 

‘‘Turbine Trip—Low Fluid Oil 

Pressure.’’ 

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 reflects a design 

change to the turbine control system that 

results in the use of an increased control oil 

pressure system, necessitating a change to the 

value at which a low fluid oil pressure 

initiates a reactor trip on turbine trip. The 

low fluid oil pressure is an input to the 

reactor trip instrumentation in response to a 

turbine trip event. The value at which the 

low fluid oil initiates a reactor trip is not an 

accident initiator. A change in the nominal 

control oil pressure does not introduce any 

mechanisms that would increase the 

probability of an accident previously 

analyzed. The reactor trip on turbine trip 

function is initiated by the same protective 

signal as used for the existing auto stop low 

fluid oil system trip signal. There is no 

change in form or function of this signal and 

the probability or consequences of previously 

analyzed accidents are not impacted. 

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 previously evaluated? 

Response: No. 

The EHC [electrohydraulic control] fluid 

oil pressure rapidly decreases in response to 

a turbine trip signal. The value at which the 

low fluid oil pressure switches initiates a 

reactor trip is not an accident initiator. The 

proposed TS change reflects the higher 

pressure that will be sensed after the pressure 

switches are relocated from the auto stop low 

fluid oil system to the EHC high pressure 

header. Failure of the new switches would 

not result in a different outcome than is 

considered in the current design basis. 

Further, the change does not alter 

assumptions made in the safety analysis but 

ensures that the instruments perform as 

assumed in the accident analysis. 

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 change involves a parameter that 

initiates an anticipatory reactor trip following 

a turbine trip. The safety analyses do not 

credit this anticipatory trip for reactor core 

protection. The original pressure switch 

configuration and the new pressure switch 

configuration both generate the same reactor 

trip signal. The difference is that the 

initiation of the trip will now be adjusted to 

a different system of higher pressure. This 

system function of sensing and transmitting 

a reactor trip signal on turbine trip remains 

the same. There is no impact to safety 

analysis acceptance criteria as described in 

the plant licensing basis because no change 

is made to the accident analysis assumptions. 

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: Sherry A. Quirk, 

Executive Vice President and General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, 6A West 

Tower, Knoxville, TN 37902. 

NRC Branch Chief: Undine Shoop. 

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 


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172 Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 

the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–003, 50–247, and 50– 

286, Indian Point Nuclear Generating 

Unit Nos. 1, 2, and 3, Westchester 

County, New York 

Date of amendment request: April 28, 

2017, as supplemented by letters dated 

August 9, 2017; September 28, 2017; 

and October 26, 2017. 

Brief description of amendments: The 

amendments revised the Cyber Security 

Plan Milestone 8 full implementation 

date by extending the full 

implementation date from December 31, 

2017, to December 31, 2018. 

Date of issuance: December 8, 2017. 

Effective date: As of the date of 

issuance, and shall be implemented by 

December 31, 2017. 

Amendment Nos.: 60 (Unit No. 1), 286 

(Unit No. 2), and 263 (Unit No. 3). A 

publicly-available version is in ADAMS 

under Accession No. ML17315A000; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Provisional Operating License No. DPR– 

5 and Facility Operating License Nos. 

DPR–26 and DPR–64: The amendments 

revised the Provisional Operating 

License for Unit No. 1 and the Facility 

Operating Licenses for Unit Nos. 2 and 

3. 

Date of initial notice in Federal 

Register: July 18, 2017 (82 FR 32880). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated December 8, 

2017. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Nuclear 

Plant, Van Buren County, Michigan 

Date of amendment request: March 

30, 2017, as supplemented by letter 

dated October 17, 2017. 

Brief description of amendment: This 

amendment revised the Cyber Security 

Plan (CSP) implementation schedule 

Milestone 8 date and paragraph 2.E in 

the renewed facility operating license 

from December 15, 2017, to March 31, 

2019. Milestone 8 of the CSP 

implementation schedule concerns the 

full implementation of the CSP. 

Date of issuance: December 15, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 264. A publiclyavailable version is in ADAMS under 

Accession No. ML17328B033; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–20: Amendment revised the 

Renewed Facility Operating License. 

Date of initial notice in Federal 

Register: May 23, 2017 (82 FR 23623). 

The supplemental letter dated October 

17, 2017, provided additional 

information that expanded the scope of 

the application as originally noticed and 

changed the NRC staff’s original 

proposed no significant hazards 

consideration (NSHC) determination as 

published in the Federal Register. 

Accordingly, the NRC published a 

second proposed NSHC determination 

in the Federal Register on November 7, 

2017 (82 FR 51650). This notice 

superseded the original notice in its 

entirety. It also provided an opportunity 

to request a hearing by January 8, 2018, 

but indicated that if the Commission 

makes a final NSHC determination, any 

such hearing would take place after 

issuance of the amendment. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 15, 

2017. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station (Pilgrim), Plymouth 

County, Massachusetts 

Date of amendment request: March 

30, 2017. 

Brief description of amendment: The 

amendment revised Pilgrim’s renewed 

facility operating license for the Cyber 

Security Plan (CSP) Milestone 8 full 

implementation completion date, as set 

forth in the CSP implementation 

schedule, and revised the physical 

protection license condition. The 

amendment revised the CSP Milestone 8 

completion date from December 15, 

2017, to December 31, 2020. 

Date of issuance: December 15, 2017. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 247. A publiclyavailable version is in ADAMS under 

Accession No. ML17290A487; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–35: The amendment revised 

the renewed facility operating license. 

Date of initial notice in Federal 

Register: May 23, 2017 (82 FR 23624). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 15, 

2017. 

No significant hazards consideration 

comments received: No. 

National Institute of Standard and 

Technology (NIST), Docket No. 50–184, 

National Bureau of Standards Test 

Reactor (NBSR), Montgomery County, 

Maryland 

Date of amendment request: March 2, 

2017, as supplemented by letters dated 

March 29, 2017; May 25, 2017; 

November 17, 2017; November 20, 2017; 

December 1, 2017; December 11, 2017; 

and December 14, 2017. 

Brief description of amendment: The 

amendment revised NIST NBSR’s 

Facility Operating License TR–5 to 

allow receipt of calibration and testing 

sources, and revised technical 

specifications pertaining to the NIST 

reactor low power startup testing and 

organizational reporting requirements. 

Date of issuance: December 15, 2017. 

Effective date: As of the date of 

issuance. 

Amendment No.: 11. A publiclyavailable version is in ADAMS under 

Accession No. ML17292A062; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. TR–5: 

Amendment revised the Renewed 

Facility Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: September 12, 2017 (82 FR 

42844). The supplemental letters dated 

November 17, 2017; November 20, 2017; 

December 1, 2017; December 11, 2017; 

and December 14, 2017 (which 

withdrew parts of the application), 

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 15, 

2017. 

No significant hazards consideration 

comments received: No. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

1 (FCS), Washington County, Nebraska 

Date of amendment request: 

December 16, 2016, as supplemented by 

letter dated May 15, 2017. 

Brief description of amendment: The 

amendment revised the FCS Emergency 

Plan and Emergency Action Level (EAL) 

scheme for the permanently defueled 

condition. The proposed permanently 

defueled Emergency Plan and EAL 

scheme are commensurate with the 


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significantly reduced spectrum of 

credible accidents that can occur in the 

permanently defueled condition and are 

necessary to properly reflect the 

conditions of the facility while 

continuing to preserve the effectiveness 

of the emergency plan. 

Date of issuance: December 12, 2017. 

Effective date: The amendment is 

effective April 7, 2018, and shall be 

implemented within 90 days of the 

effective date. 

Amendment No.: 295. A publiclyavailable version is in ADAMS under 

Accession No. ML17276B286; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–40: The amendment revised 

the Emergency Plan and EAL scheme. 

Date of initial notice in Federal 

Register: March 28, 2017 (82 FR 

15383). The supplemental letter dated 

May 15, 2017, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 12, 

2017. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: March 

27, 2017. 

Brief description of amendment: The 

licensee requested to adopt NRCapproved Technical Specifications Task 

Force (TSTF) Improved Standard 

Technical Specifications Change 

Traveler TSTF–535, Revision 0, ‘‘Revise 

Shutdown Margin Definition to Address 

Advanced Fuel Designs’’ (ADAMS 

Accession No. ML112200436), dated 

August 8, 2011. The definition of 

shutdown margin in the Hope Creek 

Generating Station Technical 

Specifications is revised to require 

calculation of shutdown margin at the 

reactor moderator temperature 

corresponding to the most reactive state 

throughout the operating cycle, which is 

68 degrees Fahrenheit or higher. 

Date of issuance: December 13, 2017. 

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. ML17317A605; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. NPF–57: Amendment revised the 

Renewed Facility Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 9, 2017 (82 FR 21560). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 13, 

2017. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: March 

27, 2017, as supplemented by letters 

dated April 28, 2017, and September 5, 

2017. 

Brief description of amendment: The 

amendment changed the Hope Creek 

Generating Station Technical 

Specifications (TSs) to relocate the 

reactor coolant system pressuretemperature (P–T) limit curves from the 

TSs to a new licensee-controlled 

document called the Pressure and 

Temperature Limits Report. The 

amendment also revised the 32 effective 

full power years P–T limit curves and 

approved P–T limit curves applicable 

through the license renewal term. The 

revisions to the curves were required 

due to the results of a recently pulled 

and tested reactor pressure vessel 

surveillance capsule. 

Date of issuance: December 14, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 209. A publiclyavailable version is in ADAMS under 

Accession No. ML17324A840; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. NPF–57: Amendment revised the 

Renewed Facility Operating License and 

TSs. 

Date of initial notice in Federal 

Register: May 23, 2017 (82 FR 23628). 

The supplemental letter dated 

September 5, 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 14, 

2017. 

No significant hazards consideration 

comments received: No. 

Southern California Edison Company, et 

al., Docket Nos. 50–206, 50–361, and 

50–362, San Onofre Nuclear Generating 

Station, Units 1, 2, and 3, San Diego 

County, California 

Date of amendment request: 

December 15, 2016, as supplemented by 

letter dated May 5, 2017. 

Brief description of amendments: The 

amendments replaced the San Onofre 

Nuclear Generating Station, Units 1, 2, 

and 3 (SONGS) Permanently Defueled 

Emergency Plan and associated 

Emergency Action Level (EAL) Bases 

Manual (hereafter referred to as the EAL 

scheme) with an Independent Spent 

Fuel Storage Installation (ISFSI) Only 

Emergency Plan (IOEP) and associated 

EAL scheme. The NRC staff determined 

that the proposed SONGS IOEP and 

associated EAL changes continue to 

meet the standards in 10 CFR 50.47, 

‘‘Emergency plans,’’ and the 

requirements in Appendix E, 

‘‘Emergency Planning and Preparedness 

for Production and Utilization 

Facilities,’’ of 10 CFR part 50, as 

exempted. As such, the SONGS IOEP 

and associated EAL changes provide 

reasonable assurance that adequate 

protective measures can and will be 

taken in the event of a radiological 

emergency. These changes more fully 

reflect the status of the facility, as well 

as the reduced scope of potential 

radiological accidents once all spent 

fuel has been moved to dry cask storage 

within the onsite ISFSI, an activity 

which is currently scheduled for 

completion in 2019. 

Date of issuance: November 30, 2017. 

Effective date: As of the date Southern 

California Edison submits a written 

notification to the NRC that all spent 

nuclear fuel assemblies have been 

transferred out of the SONGS spent fuel 

pools and placed in storage within the 

onsite ISFSI, and shall be implemented 

within 60 days. 

Amendment Nos.: 168 (Unit 1), 236 

(Unit 2), and 229 (Unit 3). A publiclyavailable version is in ADAMS under 

Accession No. ML17310B482; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Operating License Nos. DPR– 

13, NPF–10, and NPF–15: The 

amendments revised the Facility 

Operating Licenses. 

Date of initial notice in Federal 

Register: February 14, 2017 (82 FR 

10601). 

The Commission’s related evaluation 

of the amendments is contained in a 


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174 Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 

Safety Evaluation dated November 30, 

2017. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

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

Electric Generating Plant (VEGP), Units 

3 and 4, Burke County, Georgia 

Date of amendment request: May 10, 

2017, and supplemented by letter dated 

September 20, 2017. 

Description of amendments: The 

amendments consisted of changes to the 

VEGP, Units 3 and 4, Updated Final 

Safety Analysis Report in the form of 

departures from the incorporated plantspecific Design Control Document Tier 

2* and Tier 2 information (text, tables, 

and figures). Specifically, the 

amendments consisted of changes 

related to revising the design 

reinforcement in the roof of the 

auxiliary building and the design of the 

girders supporting the roof. 

Date of issuance: December 5, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment Nos.: 101 (Unit 3) and 

100 (Unit 4). A publicly-available 

version is in ADAMS under Package 

Accession No. ML17311B236; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Combined Licenses No. NPF– 

91 and NPF–92: Amendments revised 

the Facility Combined Licenses. 

Date of initial notice in Federal 

Register: June 6, 2017 (82 FR 26137). 

The supplemental letter dated 

September 20, 2017, provided 

additional information that clarified the 

application, did not expand the scope of 

the application request 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 December 5, 

2017. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

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

Electric Generating Plant (VEGP), Units 

3 and 4, Burke County, Georgia 

Date of amendment request: June 23, 

2017. 

Description of amendments: The 

amendments consisted of changes to the 

VEGP, Units 3 and 4, Updated Final 

Safety Analysis Report (UFSAR) in the 

form of departures from the plantspecific Design Control Document Tier 

2 information and involves changes to 

the VEGP, Units 3 and 4, Combined 

License Appendix A, Technical 

Specifications (TSs). Specifically, the 

proposed changes revise plant-specific 

Tier 2 information to add the time delay 

assumed in the safety analysis for the 

reactor trip on a safeguards actuation 

(‘‘S’’) signal to UFSAR Table 15.0–4a. 

This is also reflected in the proposed 

revision to TS 3.3.4, ‘‘Reactor Trip 

System (RTS) Engineered Safety Feature 

Actuation System (ESFAS) 

Instrumentation,’’ to add a surveillance 

requirement to verify the RTS response 

time for this ‘‘S’’ signal. The request also 

includes proposed changes to TS 3.3.7, 

‘‘RTS Trip Actuation Devices,’’ to clarify 

that the requirements for reactor trip 

breaker (RTB) undervoltage and shunt 

trip mechanisms apply only to inservice RTBs. In addition, the request 

includes proposed changes to TS 3.3.9, 

‘‘ESFAS Manual Initiation,’’ to correct 

the nomenclature for the Chemical and 

Volume Control System, which is 

inadvertently stated as the Chemical 

Volume and Control System. 

Date of issuance: December 8, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment Nos.: 102 (Unit 3) and 

101 (Unit 4). A publicly-available 

version is in ADAMS under Accession 

No. ML17296A236; documents related 

to these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Facility Combined Licenses No. NPF– 

91 and NPF–92: Amendments revised 

the Facility Combined Licenses. 

Date of initial notice in Federal 

Register: August 15, 2017 (82 FR 

38714). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated December 8, 

2017. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

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

Electric Generating Plant (VEGP), Units 

3 and 4, Burke County, Georgia 

Date of amendment request: October 

20, 2016. 

Description of amendments: The 

amendments authorized changes to the 

Tier 2* information in the VEGP, Units 

3 and 4, Updated Final Safety Analysis 

Report (which includes the plantspecific design control document 

information) to clarify the 

demonstration of the quality and 

strength of a specific set of couplers 

welded to carbon steel embedment 

plates, already installed and embedded 

in concrete through visual examination 

and static tension testing, in lieu of the 

nondestructive examination 

requirements of American Institute of 

Steel Construction (AISC) N690. 

Date of issuance: September 5, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment Nos.: 86 (Unit 3) and 85 

(Unit 4). A publicly-available version is 

in ADAMS under Package Accession 

No. ML17178A197; documents related 

to these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Facility Combined Licenses Nos. NPF– 

91 and NPF–92: Amendments revised 

the Facility Combined Licenses. 

Date of initial notice in Federal 

Register: March 14, 2017 (82 FR 

13662). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 5, 

2017. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of amendment request: March 

31, 2017. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 5.7.2.14, ‘‘Ventilation 

Filter Testing Program (VFTP),’’ to 

correct an administrative error 

introduced by Amendment No. 92, 

issued June 19, 2013. Specifically, 

Amendment 92 deleted TS 3.9.8, 

‘‘Reactor Building Purge Air Cleanup 

Units,’’ but did not delete associated 

references to the reactor building purge 

filters from TS 5.7.2.14. 

Date of issuance: December 7, 2017. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 117. A publiclyavailable version is in ADAMS under 

Accession No. ML17311A786; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. NPF– 

90: Amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 5, 2017 (82 FR 31103). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 7, 

2017. 

No significant hazards consideration 

comments received: No. 


daltland on DSKBBV9HB2PROD with NOTICES
Federal Register / Vol. 83, No. 1 / Tuesday, January 2, 2018 / Notices 175 

Dated at Rockville, Maryland, this 21st day 

of December 2017. 

For the Nuclear Regulatory Commission. 

Kathryn M. Brock, 

Acting Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2017–27930 Filed 12–29–17; 8:45 am] 

BILLING CODE 7590–01–P 

POSTAL REGULATORY COMMISSION 

[Docket Nos. CP2017–177, MC2018–76 and 

CP2018–118; MC2018–77 and CP2018–119; 

MC2018–78 and CP2018–120; MC2018–79 

and CP2018–121; MC2018–80 and CP2018– 

122; MC2018–81 and CP128–123; MC2018– 

82 and CP2018–124; MC2018–83 and 

CP2018–125; MC2018–84 and CP2018–126; 

MC2018–85 and CP2018–127; MC2018–86 

and CP2018–128] 

New Postal Products 

AGENCY: Postal Regulatory Commission. 

ACTION: Notice. 

SUMMARY: The Commission is noticing 

recent Postal Service filings for the 

Commission’s consideration concerning 

negotiated service agreements. This 

notice informs the public of the filing, 

invites public comment, and takes other 

administrative steps. 

DATES: Comments are due: January 3, 

2018, January 4, 2018, and January 5, 

2018. 

ADDRESSES: Submit comments 

electronically via the Commission’s 

Filing Online system at http://

www.prc.gov. Those who cannot submit 

comments electronically should contact 

the person identified in the FOR FURTHER

INFORMATION CONTACT section by 

telephone for advice on filing 

alternatives. 

FOR FURTHER INFORMATION CONTACT: 

David A. Trissell, General Counsel, at 

202–789–6820. 

SUPPLEMENTARY INFORMATION: The 

January 3, 2018 comment due date 

applies to Docket Nos. MC2018–76 and 

CP2018–118; MC2018–77 and CP2018– 

119; MC2018–78 and CP2018–120; 

MC2018–79 and CP2018–121; MC2018– 

80 and CP2018–122. 

The January 4, 2018 comment due 

date applies to Docket Nos. MC2018–81 

and CP2018–123; MC2018–82 and 

CP2018–124; MC2018–83 and CP2018– 

125; MC2018–84 and CP2018–126; 

MC2018–85 and CP2018–127. 

The January 5, 2018 commend due 

date applies to Docket Nos. CP2017– 

177; MC2018–86 and CP2018–128. 

Table of Contents 

I. Introduction 

II. Docketed Proceeding(s) 

I. Introduction 

The Commission gives notice that the 

Postal Service filed request(s) for the 

Commission to consider matters related 

to negotiated service agreement(s). The 

request(s) may propose the addition or 

removal of a negotiated service 

agreement from the market dominant or 

the competitive product list, or the 

modification of an existing product 

currently appearing on the market 

dominant or the competitive product 

list. 

Section II identifies the docket 

number(s) associated with each Postal 

Service request, the title of each Postal 

Service request, the request’s acceptance 

date, and the authority cited by the 

Postal Service for each request. For each 

request, the Commission appoints an 

officer of the Commission to represent 

the interests of the general public in the 

proceeding, pursuant to 39 U.S.C. 505 

(Public Representative). Section II also 

establishes comment deadline(s) 

pertaining to each request. 

The public portions of the Postal 

Service’s request(s) can be accessed via 

the Commission’s website (http://

www.prc.gov). Non-public portions of 

the Postal Service’s request(s), if any, 

can be accessed through compliance 

with the requirements of 39 CFR 

3007.40. 

The Commission invites comments on 

whether the Postal Service’s request(s) 

in the captioned docket(s) are consistent 

with the policies of title 39. For 

request(s) that the Postal Service states 

concern market dominant product(s), 

applicable statutory and regulatory 

requirements include 39 U.S.C. 3622, 39 

U.S.C. 3642, 39 CFR part 3010, and 39 

CFR part 3020, subpart B. For request(s) 

that the Postal Service states concern 

competitive product(s), applicable 

statutory and regulatory requirements 

include 39 U.S.C. 3632, 39 U.S.C. 3633, 

39 U.S.C. 3642, 39 CFR part 3015, and 

39 CFR part 3020, subpart B. Comment 

deadline(s) for each request appear in 

section II. 

II. Docketed Proceeding(s) 

1. Docket No(s).: CP2017–177; Filing 

Title: USPS Notice of Change in Prices 

Pursuant to Amendment to Priority Mail 

Express, Priority Mail & First-Class 

Package Service Contract 17; Filing 

Acceptance Date: December 21, 2017; 

Filing Authority: 39 CFR 3015.5; Public 

Representative: Gregory Stanton; 

Comments Due: January 5, 2018. 

2. Docket No(s).: MC2018–76 and 

CP2018–118; Filing Title: USPS Request 

to Add Priority Mail Express, Priority 

Mail & First-Class Package Service 

Contract 31 to Competitive Product List 

and Notice of Filing Materials Under 

Seal; Filing Acceptance Date: December 

21, 2017; Filing Authority: 39 U.S.C. 

3642 and 39 CFR 3020.30 et seq.; Public 

Representative: Gregory Stanton; 

Comments Due: January 3, 2018. 

3. Docket No(s).: MC2018–77 and 

CP2018–119; Filing Title: USPS Request 

to Add Priority Mail Contract 401 to 

Competitive Product List and Notice of 

Filing Materials Under Seal; Filing 

Acceptance Date: December 21, 2017; 

Filing Authority: 39 U.S.C. 3642 and 39 

CFR 3020.30 et seq.; Public 

Representative: Michael L. Leibert; 

Comments Due: January 3, 2018. 

4. Docket No(s).: MC2018–78 and 

CP2018–120; Filing Title: USPS Request 

to Add First-Class Package Service 

Contract 89 to Competitive Product List 

and Notice of Filing Materials Under 

Seal; Filing Acceptance Date: December 

21, 2017; Filing Authority: 39 U.S.C. 

3642 and 39 CFR 3020.30 et seq.; Public 

Representative: Curtis E. Kidd; 

Comments Due: January 3, 2018. 

5. Docket No(s).: MC2018–79 and 

CP2018–121; Filing Title: USPS Request 

to Add First-Class Package Service 

Contract 90 to Competitive Product List 

and Notice of Filing Materials Under 

Seal; Filing Acceptance Date: December 

21, 2017; Filing Authority: 39 U.S.C. 

3642 and 39 CFR 3020.30 et seq.; Public 

Representative: Curtis E. Kidd; 

Comments Due: January 3, 2018. 

6. Docket No(s).: MC2018–80 and 

CP2018–122; Filing Title: USPS Request 

to Add Priority Mail Contract 402 to 

Competitive Product List and Notice of 

Filing Materials Under Seal; Filing 

Acceptance Date: December 21, 2017; 

Filing Authority: 39 U.S.C. 3642 and 39 

CFR 3020.30 et seq.; Public 

Representative: Michael L. Leibert; 

Comments Due: January 3, 2018. 

7. Docket No(s).: MC2018–81 and 

CP2018–123; Filing Title: USPS Request 

to Add Priority Mail Express & Priority 

Mail Contract 55 to Competitive Product 

List and Notice of Filing Materials 

Under Seal; Filing Acceptance Date: 

December 21, 2017; Filing Authority: 39 

U.S.C. 3642 and 39 CFR 3020.30 et seq.; 

Public Representative: Matthew R. 

Ashford; Comments Due: January 4, 

2018. 

8. Docket No(s).: MC2018–82 and 

CP2018–124; Filing Title: USPS Request 

to Add Priority Mail & First-Class 

Package Service Contract 67 to 

Competitive Product List and Notice of 

Filing Materials Under Seal; Filing 

Acceptance Date: December 22, 2017; 

Filing Authority: 39 U.S.C. 3642 and 39 

CFR 3020.30 et seq.; Public 

Representative: Timothy J. Schwuchow; 

Comments Due: January 4, 2018. 


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