Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17083 

The Commission has determined for 

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 on 

October 14, 2014 (79 FR 61662). The 

September 23, 2014, application 

revision, and the October 30 and 

November 6, 2014, supplements had no 

effect on the no significant hazards 

consideration determination, and no 

comments were received during the 60- 

day comment period. 

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. 

IV. Conclusion 

Using the reasons set forth in the 

combined safety evaluation, the staff 

granted the exemption and issued the 

amendment that the licensee requested 

on August 22, 2014, and revised by 

letter dated September 23, 2014, and 

supplemented by letters dated October 

30 and November 6, 2014. The 

exemption and amendment were issued 

on December 23, 2014 as part of a 

combined package to the licensee 

(ADAMS Accession No. ML14323A609). 

Dated at Rockville, Maryland, this 23rd day 

of March 2015. 

For the Nuclear Regulatory Commission. 

Lawrence Burkhart, 

Chief, Licensing Branch 4, Division of New 

Reactor Licensing, Office of New Reactors. 

[FR Doc. 2015–07277 Filed 3–30–15; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2015–0073] 

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 March 5, 

2015 to March 18, 2015. The last 

biweekly notice was published on 

March 17, 2015. 

DATES: Comments must be filed by April 

30, 2015. A request for a hearing must 

be filed by June 1, 2015. 

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 Web site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2015–0073. 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: Cindy Bladey, 

Office of Administration, Mail Stop: 

OWFN–12–H08, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001. 

For additional direction on obtaining 

information and submitting comments, 

see ‘‘Obtaining Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: Kay 

Goldstein, Office of Nuclear Reactor 

Regulation, U.S. Nuclear Regulatory 

Commission, Washington DC 20555– 

0001; telephone: 301–415–1506, email: 

Kay.Goldstein@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2015– 

0073 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 Web site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2015–0073. 

• 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 the SUPPLEMENTARY

INFORMATION section. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

B. Submitting Comments 

Please include Docket ID NRC–2015– 

0073, facility name, unit number(s), 

application date, and subject in your 

comment submission. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 

disclosed in your comment submission. 

The NRC posts all comment 

submissions at http://

www.regulations.gov as well as entering 

the comment submissions into ADAMS. 

The NRC does not routinely edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information that 

they do not want to be publicly 

disclosed in their comment submission. 

Your request should state that the NRC 

does not routinely edit comment 


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submissions to remove such information 

before making the comment 

submissions available to the public or 

entering the comment submissions into 

ADAMS. 

II. Notice of Consideration of Issuance 

of Amendments to Facility Operating 

Licenses and Combined Licenses and 

Proposed No Significant Hazards 

Consideration Determination 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

§ 50.92 of Title 10 of the Code of Federal 

Regulations (10 CFR), this means that 

operation of the facility in accordance 

with the proposed amendment would 

not (1) involve a significant increase in 

the probability or consequences of an 

accident previously evaluated, or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

A. Opportunity To Request a Hearing 

and Petition for Leave To Intervene 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license or 

combined license. Requests for a 

hearing and a petition for leave to 

intervene shall be filed in accordance 

with the Commission’s ‘‘Agency Rules 

of Practice and Procedure’’ in 10 CFR 

part 2. Interested person(s) should 

consult a current copy of 10 CFR 2.309, 

which is available at the NRC’s PDR, 

located at One White Flint North, Room 

O1–F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland 20852. The 

NRC’s regulations are accessible 

electronically from the NRC Library on 

the NRC’s Web site at http://

www.nrc.gov/reading-rm/doccollections/cfr/. If a request for a hearing 

or petition for leave to intervene is filed 

by the above date, the Commission or a 

presiding officer designated by the 

Commission or by the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board Panel, will 

rule on the request and/or petition; and 

the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/

petitioner to relief. A requestor/

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, then any hearing held 

would take place before the issuance of 

any amendment unless the Commission 

finds an imminent danger to the health 

or safety of the public, in which case it 

will issue an appropriate order or rule 

under 10 CFR part 2. 

B. Electronic Submissions (E-Filing) 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, 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 participating 

under 10 CFR 2.315(c), must be filed in 

accordance with the NRC’s E-Filing rule 

(72 FR 49139; August 28, 2007). The EFiling process requires participants to 

submit and serve all adjudicatory 


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documents over the internet, or in some 

cases to mail copies on electronic 

storage media. 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 ten 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 request (1) a digital 

identification (ID) certificate, which 

allows the participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

representative, already holds an NRCissued digital ID certificate). Based upon 

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on the 

NRC’s public Web site at http://

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

getting-started.html. System 

requirements for accessing the ESubmittal server are detailed in the 

NRC’s ‘‘Guidance for Electronic 

Submission,’’ which is available on the 

agency’s public Web site at http://

www.nrc.gov/site-help/esubmittals.html. Participants may 

attempt to use other software not listed 

on the Web site, but should note that the 

NRC’s E-Filing system does not support 

unlisted software, and the NRC Meta 

System Help Desk will not be able to 

offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through the Electronic 

Information Exchange System, users 

will be required to install a Web 

browser plug-in from the NRC’s Web 

site. Further information on the Webbased submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

site at http://www.nrc.gov/site-help/esubmittals.html. 

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC’s public Web site 

at http://www.nrc.gov/site-help/esubmittals.html. A filing is considered 

complete at the time the documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

system no later than 11:59 p.m. Eastern 

Time on the due date. Upon receipt of 

a transmission, the E-Filing system 

time-stamps the document and sends 

the submitter an email notice 

confirming receipt of the document. The 

E-Filing system also distributes an email 

notice that provides access to the 

document to the NRC’s Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the NRC’s adjudicatory E-Filing system 

may seek assistance by contacting the 

NRC Meta System Help Desk through 

the ‘‘Contact Us’’ link located on the 

NRC’s public Web site 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 

Meta System Help Desk is available 

between 8 a.m. and 8 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 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, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

Filing is considered complete by firstclass mail as of the time of deposit in 

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. A presiding 

officer, having granted an exemption 

request from using E-Filing, may require 

a participant or party to use E-Filing if 

the presiding officer subsequently 

determines that the reason for granting 

the exemption from use of E-Filing no 

longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in the NRC’s 

electronic hearing docket which is 

available to the public at http://

ehd1.nrc.gov/ehd/, unless excluded 

pursuant to an order of the Commission, 

or the presiding officer. Participants are 

requested not to include personal 

privacy information, such as social 

security numbers, home addresses, or 

home phone numbers in their filings, 

unless an NRC regulation or other law 

requires submission of such 

information. However, a request to 

intervene will require including 

information on local residence in order 

to demonstrate a proximity assertion of 

interest in the proceeding. 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. 

Petitions for leave to intervene must 

be filed no later than 60 days from the 

date of publication of this notice. 

Requests for hearing, petitions for leave 

to intervene, and motions for leave to 

file new or amended contentions that 

are filed after the 60-day deadline will 

not be entertained absent a 

determination by the presiding officer 

that the filing demonstrates good cause 

by satisfying the three factors in 10 CFR 

2.309(c)(1)(i)–(iii). 

For further details with respect to 

these license amendment applications, 

see the application for amendment 

which is available for public inspection 

in ADAMS and at the NRC’s PDR. For 

additional direction on accessing 

information related to this document, 

see the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–413 and 50–414, Catawba 

Nuclear Station, Units 1 and 2, York 

County, South Carolina 

Date of amendment request: 

November 24, 2014. A publiclyVerDate Sep<11>2014 18:32 Mar 30, 2015 Jkt 235001 PO 00000 Frm 00061 Fmt 4703 Sfmt 4703 E:\FR\FM\31MRN1.SGM 31MRN1

asabaliauskas on DSK5VPTVN1PROD with NOTICES
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available version is in ADAMS under 

Accession No. ML14330A327. 

Description of amendment request: 

The proposed amendments would 

modify the Technical Specifications 

(TS) to correct non-conservative 

setpoints. Specifically, modify the 

Allowable Value parameter and the 

Nominal Trip Setpoint for the TS 3.3.2 

Table 3.3.2–1, ‘‘Engineered Safety 

Feature Actuation System 

Instrumentation’’ function for Auxiliary 

Feedwater Loss of Offsite Power 

(Function 6.d.) and for the TS 3.3.5 Loss 

of Voltage function in Surveillance 

Requirement (SR) 3.3.5.2. As part of the 

change, the licensee is also proposing to 

add the applicable footnotes in 

accordance with TSTF–493, Revision 4, 

‘‘Clarify Application of Setpoint 

Methodology for LSSS [limiting safety 

system set point] Functions.’’ 

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 

and staff’s changes/additions are 

provided in [ ]: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Duke Energy requests NRC review and 

approval to revise the Allowable Value 

parameter for the Technical Specification 

(TS) 3.3.2 Table 3.3.2–1, ‘‘Engineered Safety 

Feature Actuation System Instrumentation’’ 

function for Auxiliary Feedwater Loss of 

Offsite Power (Function 6.d.) and for the TS 

3.3.5 Loss of Voltage function in Surveillance 

Requirement (SR) 3.3.5.2 in order to make 

this parameter more restrictive. The existing 

parameter was determined to be nonconservative and this parameter is presently 

classified as Operable But Degraded in the 

Catawba Corrective Action Program. In 

addition, the Nominal Trip Setpoint 

parameter for this function is being slightly 

lowered in order to gain additional margin. 

Finally, as part of this License Amendment 

Request (LAR), applicable footnotes are also 

being added to the affected TS 3.3.2 function 

in accordance with TS Task Force Traveler 

[(TSTF)] TSTF–493, Revision 4, ‘‘Clarify 

Application of Setpoint Methodology for 

LSSS Functions.’’ The more restrictive 

Allowable Value will preclude the potential 

for a double sequencing event to occur under 

the condition of a Loss of Coolant Accident 

(LOCA) load sequencer actuation with a preexisting degraded voltage condition on the 

essential buses. These proposed changes will 

not increase the probability of occurrence of 

any design basis accident since the affected 

function, in and of itself, cannot initiate an 

accident. Should a LOCA occur, the 

proposed changes will ensure that the 

sequencer operates properly in order to 

mitigate the consequences of the event. 

Appropriate calculations were developed to 

substantiate the revised TS parameters 

proposed in this LAR. There will be no 

impact on the source term or pathways 

assumed in accidents previously evaluated. 

No analysis assumptions will be violated and 

there will be no adverse effects on onsite or 

offsite doses as the result of an accident. 

Adoption of the TSTF–493 footnotes for the 

respective SRs will ensure that the function’s 

channels will continue to behave in 

accordance with safety analysis assumptions 

and the channel performance assumptions in 

the setpoint methodology. 

Therefore, the proposed amendments 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 amendments do not change 

the methods governing normal plant 

operation; nor are the methods utilized to 

respond to plant transients altered. In 

addition, the proposed changes to the 

affected TS parameters and the adoption of 

the TSTF–493 footnotes will not create the 

potential for any new initiating events or 

transients to occur in the actual physical 

plant. 

Therefore, the proposed amendments do 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

Response: No. 

Margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their design 

functions during and following an accident. 

These barriers include the fuel cladding, the 

reactor coolant system, and the containment 

system. The proposed changes will assure the 

acceptable operation of the affected function 

under all postulated transient and accident 

conditions. This will ensure that all 

applicable design and safety limits are 

satisfied such that the fission product 

barriers will continue to perform their design 

functions. 

Therefore, the proposed amendments do 

not involve a significant reduction in a 

margin of safety. 

Based on the preceding discussion, Duke 

Energy concludes that the proposed 

amendments do not involve a significant 

hazards consideration under the standards 

set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lara S. Nichols, 

Associate General Counsel, Duke Energy 

Corporation, 526 South Church Street— 

EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: March 

14, 2014. A publicly-available version is 

in ADAMS under Accession No. 

ML14078A037. 

Description of amendment request: 

The amendment would revise the 

Technical Specifications (TS) for the 

Inservice Testing Program to reflect the 

current edition of the American Society 

of Mechanical Engineers (ASME) Code 

that is referenced in 10 CFR 50.55a(b). 

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 corrects a 

typographical error in TS 5.5.8, ‘‘Reactor 

Coolant Pump Flywheel Inspection 

Program,’’ and revises TS 5.5.9, ‘‘lnservice 

Testing Program,’’ for consistency with the 

requirements of 10 CFR 50.55a(f)(4) regarding 

the inservice testing of pumps and valves 

which are classified as ASME Code Class 1, 

Class 2 and Class 3. The proposed change 

incorporates revisions to the ASME Code that 

result in a net improvement in the measures 

for testing pumps and valves. 

The proposed change does not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. The proposed change does not 

involve the addition or removal of any 

equipment, or any design changes to the 

facility. 

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 corrects a 

typographical error in TS 5.5.8, ‘‘Reactor 

Coolant Pump Flywheel Inspection 

Program,’’ and revises TS 5.5.9, ‘‘lnservice 

Testing Program,’’ for consistency with the 

requirements of 10 CFR 50.55a(f)(4) regarding 

the inservice testing of pumps and valves 

which are classified as ASME Code Class 1, 

Class 2 and Class 3. The proposed change 

incorporates revisions to the ASME Code that 

result in a net improvement in the measures 

for testing pumps and valves. 

The proposed change does not involve a 

modification to the physical configuration of 


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the plant (i.e., no new equipment will be 

installed), nor does it involve a change in the 

methods governing normal plant operation. 

The proposed change will not impose any 

new or different requirements or introduce a 

new accident initiator, accident precursor, or 

malfunction mechanism. Additionally, there 

is no change in the types or increases in the 

amounts of any effluent that may be released 

offsite and there is no increase in individual 

or cumulative occupational exposure. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change corrects a 

typographical error in TS 5.5.8, ‘‘Reactor 

Coolant Pump Flywheel Inspection 

Program,’’ and revises TS 5.5.9, ‘‘lnservice 

Testing Program,’’ for consistency with the 

requirements of 10 CFR 50.55a(f)(4) regarding 

the inservice testing of pumps and valves 

which are classified as ASME Code Class 1, 

Class 2 and Class 3. The proposed change 

incorporates revisions to the ASME Code that 

result in a net improvement in the measures 

for testing pumps and valves. The safety 

function of the affected pumps and valves 

will be maintained. Therefore, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lara S. Nichols, 

Deputy General Counsel, Duke Energy 

Corporation, 550 South Tryon Street— 

DEC45A, Charlotte, NC 28202–1802. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc., Docket No. 50–416, 

Grand Gulf Nuclear Station, Unit 1 

(GGNS), Claiborne County, Mississippi 

Date of amendment request: 

November 21, 2014. A publiclyavailable version is in ADAMS under 

Accession No. ML14325A520. 

Description of amendment request: 

The amendment would change the 

GGNS Technical Specification (TS) 

2.1.1, ‘‘Reactor Core SLs [Safety 

Limits].’’ Specifically, the change would 

revise the Minimum Critical Power 

Ratio (MCPR) SL stated in TS 2.1.1.2 for 

two-loop operation from greater than or 

equal to (≥) 1.11 to ≥ 1.15. Additionally, 

the change would revise the MCPR SL 

stated in TS 2.1.1.2 for single-loop 

operation from ≥ 1.14 to ≥ 1.15. 

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 Bases to TS 2.1.1.2 states that: ‘‘The 

MCPR SL ensures sufficient conservatism in 

the operating MCPR limit that, in the event 

of an AOO [Anticipated Operational 

Occurrence] from the limiting condition of 

operation, at least 99.9% of the fuel rods in 

the core would be expected to avoid boiling 

transition. 

This condition is met in that the GGNS 

Cycle 20 (C20) MCPR SL evaluation was 

performed in accordance with Reference 4 

[NEDE–24011–P–A, ‘‘General Electric 

Standard Application for Reactor Fuel 

(GESTAR–II’’)]. The resulting values 

continue to ensure the conservatism 

described in the Bases to TS 2.1.1.2. The 

proposed changes also continue to ensure 

sufficient conservatism in the operating 

MCPR limit. The MCPR operating limits are 

presented and controlled in accordance with 

the GGNS Core Operating Limits Report 

(COLR). 

The requested Technical Specification 

change does not involve any plant 

modifications or operational changes that 

could affect system reliability or performance 

or that could affect the probability of operator 

error. The requested change does not affect 

any postulated accident precursors, any 

accident mitigating systems, or introduce any 

new accident initiation mechanisms. 

Therefore, the proposed change to increase 

the MCPR SL values 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 involve any 

new modes of operation, any changes to 

setpoints, or any plant modifications. The 

proposed change to the MCPR SL accounts 

for requirements specified in the NRC Safety 

Evaluation limitations and conditions 

associated with NEDC–33173P 

[‘‘Applicability of GE Methods to Expanded 

Operating Domains’’] and NEDC–33006P 

[‘‘Licensing Topical Report—General Electric 

Boiling Water Reactor Maximum Extended 

Load Line Limit Analysis Plus’’]. Compliance 

with the criterion for incipient boiling 

transition continues to be ensured. The core 

operating limits will continue to be 

developed using NRC approved methods. 

The proposed [MCPR SL] does not result in 

the creation of any new precursors to an 

accident. 

Therefore, the proposed change does not 

create 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 MCPR SLs have been evaluated in 

accordance with Global Nuclear Fuels NRCapproved cycle-specific safety limit 

methodology to ensure that during normal 

operation and during AOO’s, at least 99.9% 

of the fuel rods in the core are not expected 

to experience transition boiling. The 

proposed change to the [MCPR SL] accounts 

for requirements specified in the NRC Safety 

Evaluation limitations and conditions 

associated with NEDC–33173P and NEDC– 

33006P, which result in additional margin 

above that specified in the TS Bases. 

Therefore, the proposed change to the 

MCPR SL does not involve a significant 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Meena K. Khanna. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc., Docket No. 50–416, 

Grand Gulf Nuclear Station, Unit 1 

(GGNS), Claiborne County, Mississippi 

Date of amendment request: 

November 21, 2014, as supplemented by 

letter dated February 18, 2015. Publiclyavailable versions are in ADAMS under 

Accession Nos. ML14325A752 and 

ML15049A536, respectively. 

Description of amendment request: 

The proposed amendment would revise 

GGNS’s license basis to adopt a single 

fluence methodology. 

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 adopts a single flux 

methodology. While Chapter 15, Accident 

Analysis, of the Standard Review Plan 

(NUREG–0800, Standard Review Plan for the 

Review of Safety Analysis Reports for 

Nuclear Power Plants) assumes the pressure 

vessel does not fail, the flux methodology is 

not an initiator to any accident previously 

evaluated. Accordingly, the proposed change 


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to the adoption of the flux methodology has 

no effect on the probability of any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change adopts a flux 

methodology. The change does not involve a 

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

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operations. The 

change does not alter assumptions made in 

the safety analysis regarding fluence. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change adopts a single 

fluence methodology. The proposed change 

does not alter the manner in which safety 

limits, limiting safety system settings or 

limiting conditions for operation are 

determined. The proposed change ensures 

that the methodology used for fluence is in 

compliance with RG 1.190 requirements. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Meena K. Khanna. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois 

Date of amendment request: August 

19, 2014. A publicly-available version is 

in ADAMS under Accession No. 

ML14231A902. 

Description of amendment request: 

The proposed amendment would 

increase the technical specification (TS) 

surveillance requirement (SR) 3.7.9.2 

allowable temperature to less than or 

equal to 102 °F [degree Fahrenheit]. 

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 likelihood of a malfunction of any 

systems, structures or components (SSCs) 

supported by the UHS [ultimate heat sink] is 

not significantly increased by increasing the 

allowable Ultimate Heat Sink (UHS) 

temperature from ≤100 °F to ≤102 °F. The 

UHS provides a heat sink for process and 

operating heat from safety related 

components during a transient or accident, as 

well as during normal operation. The 

proposed change does not make any physical 

changes to any plant SSCs, nor does it alter 

any of the assumptions or conditions upon 

which the UHS is designed. The UHS is not 

an initiator of any analyzed accident. All 

equipment supported by the UHS has been 

evaluated to demonstrate that their 

performance and operation remains as 

described in the UFSAR [updated final safety 

analysis report] with no increase in 

probability of failure or malfunction. 

The SSCs credited to mitigate the 

consequences of postulated design basis 

accidents remain capable of performing their 

design basis function. The change in 

maximum UHS temperature has been 

evaluated using the UFSAR described 

methods to demonstrate that the UHS 

remains capable of removing normal 

operating and post-accident heat. The change 

in UHS temperature and resulting 

containment response following a postulated 

design basis accident has been demonstrated 

to not be impacted. Additionally, all the UHS 

supported equipment, credited in the 

accident analysis to mitigate an accident, has 

been shown to continue to perform their 

design function as described in the UFSAR. 

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 accident previously 

evaluated. The proposed change does not 

introduce any new modes of plant operation, 

change the design function of any SSC, 

change the mode of operation of any SSC, or 

change any actions required when the TS 

limit is exceeded. There are no new 

equipment failure modes or malfunctions 

created as affected SSCs continue to operate 

in the same manner as previously evaluated 

and have been evaluated to perform as 

designed at the increased UHS temperature 

and as assumed in the accident analysis. 

Additionally, accident initiators remain as 

described in the UFSAR and no new accident 

initiators are postulated as a result of the 

increase in UHS temperature. 

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 change continues to ensure 

that the maximum temperature of the cooling 

water supplied to the plant SSCs during a 

UHS design basis event remains within the 

evaluated equipment limits and capabilities 

assumed in the accident analysis. The 

proposed change does not result in any 

changes to plant equipment function, 

including setpoints and actuations. All 

equipment will function as designed in the 

plant safety analysis without any physical 

modifications. The proposed change does not 

alter a limiting condition for operation, 

limiting safety system setting, or safety limit 

specified in the Technical Specifications. 

The proposed change does not adversely 

impact the UHS inventory required to be 

available for the UFSAR described design 

basis accident involving the worst case 30- 

day period including losses for evaporation 

and seepage to support safe shutdown and 

cooldown of both Braidwood Station units. 

Additionally, the structural integrity of the 

UHS is not impacted and remains acceptable 

following the change, thereby ensuring that 

the assumptions for both UHS temperature 

and inventory remain valid. 

Therefore, since there is no adverse impact 

of this change on the Braidwood Station 

safety analysis, there is no reduction in the 

margin of safety of the plant. 

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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Travis L. Tate 

Exelon Generation Company, LLC 

(EGC), Docket Nos. STN 50–454 and 

STN 50–455, Byron Station, Units 1 and 

2, Ogle County, Illinois 

Date of amendment request: 

November 24, 2014. A publiclyavailable version is in ADAMS under 

Accession No. ML14328A800. 

Description of amendment request: 

The proposed amendment would revise 

Condition I and surveillance 

requirement (SR) 3.7.9.3 associated with 

technical specification (TS) Section 

3.7.9, ‘‘Ultimate Heat Sink (UHS),’’ to 

reflect the current design basis flood 

level. 

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 


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consideration, which is presented 

below: 

EGC has evaluated whether or not a 

significant hazards consideration is involved 

with the proposed amendment by focusing 

on the three standards set forth in 10 CFR 

50.92(c), ‘‘Issuance of amendment,’’ as 

discussed below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to revise TS 3.7.9, 

Condition I and SR 3.7.9.3 will ensure the 

operability of the SX [service water] makeup 

pumps to meet TS 3.7.9 LCO [Limiting 

Condition for Operation] requirement. The 

proposed change does not result in any 

physical changes to safety related structures, 

systems, or components. The probability of a 

flood at the river screen house (RSH) is 

unchanged. Since the UHS itself is not an 

accident initiator, the proposed change does 

not impact the initiators or assumptions of 

analyzed accidents, nor do they impact the 

mitigation of accidents or transient events. 

Consequently, the proposed change does not 

increase the probability of occurrence for any 

accident previously evaluated. 

The proposed change will ensure that 

actions to verify operability of the deep well 

pumps will be taken prior to the potential for 

the SX makeup pumps to be adversely 

affected by the combined event flood high 

river level. Therefore, the UHS will be 

capable of performing its functions to 

mitigate accidents by serving as the heat sink 

for safety related equipment. Thus, the 

proposed change does not increase the 

consequences of any accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to revise TS 3.7.9, 

Condition I and SR 3.7.9.3 does not change 

the design function or operation of the SX 

makeup pumps. The proposed change does 

not change or introduce the possibility of any 

new or different type of equipment, modes of 

system operation, failure mechanisms, 

malfunctions, or accident initiators. The 

proposed change to lower the river level 

value at which action is taken to verify basin 

levels and deep well pumps are ready to 

perform the UHS makeup function in the 

place of the SX makeup pumps will not affect 

the operation or function of the UHS or the 

deep well pumps. 

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. 

The proposed change to revise TS 3.7.9, 

Condition I and SR 3.7.9.3 reestablishes the 

margin between the design bases combined 

event flood level and TS 3.7.9, Condition I 

action level for high river level. The 

proposed change will ensure the operability 

of the SX makeup pumps to meet TS 3.7.9 

LCO and do not affect the ability of the SX 

makeup pumps to provide the safety related 

source makeup to the UHS. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, EGC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and 

accordingly, a finding of no significant 

hazards consideration is justified. 

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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Travis L. Tate. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Exelon Generation Company, LLC, 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of amendment request: 

December 22, 2014. A publicly-available 

version is in ADAMS under Accession 

No. ML14357A085. 

Description of amendment request: 

The proposed amendment modifies the 

technical specifications (TSs) to add a 

new Limiting Condition for Operation 

(LCO) 3.10.8 to specifically permit 

inservice leakage and hydrostatic testing 

at reactor coolant system (RCS) 

temperatures greater than the average 

reactor coolant temperature for MODE 4 

with the reactor shutdown. In addition, 

the proposed amendment includes an 

expanded scope of LCO 3.10.8 

consistent with the NRC-approved 

Revision 0 of Technical Specification 

Task Force (TSTF) Improved Standard 

Technical Specification Change 

Traveler, TSTF–484, ‘‘Use of TS 3.10.1 

for Scram Time Testing Activities’’ 

available in ADAMS under Accession 

No. ML062990425. 

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: 

EGC [Exelon Generation Company] has 

evaluated the proposed changes, using the 

criteria in 10 CFR 50.92, and has determined 

that the proposed changes do not involve a 

significant hazards consideration. The 

following information is provided to support 

a finding of no significant hazards 

consideration. 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes will not result in a 

significant change in the stored energy in the 

reactor vessel during the performance of the 

testing. The probability of an accident is not 

significantly increased because the proposed 

changes will not alter the method by which 

inservice leakage and hydrostatic testing is 

performed or significantly change the 

temperatures and pressures achieved to 

perform the test. 

The consequences of previously evaluated 

accidents are not significantly increased 

because the required testing conditions 

provide adequate assurance that the 

consequences of a steam leak will be 

conservatively bounded by the consequences 

of the postulated main system line break 

outside of primary containment. Under these 

proposed changes, the secondary 

containment, standby gas treatment system, 

and associated initiation instrumentation are 

required to be operable during the 

performance of inservice leakage and 

hydrostatic testing and would be capable of 

mitigating any airborne radioactivity or steam 

leaks that could occur. In addition, the 

required Emergency Core Cooling subsystems 

will be more than adequate to ensure that a 

significant increase in consequences will not 

occur by ensuring that the potential for failed 

fuel and a subsequent increase in coolant 

activity above Technical Specification limits 

are minimized. 

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. 

As the accumulated neutron fluence on the 

reactor vessel increases, the PressureTemperature Limits in TS 3.4.9 for DNPS 

[Dresden Nuclear Power Station] and QCNPS 

[Quad Cities Nuclear Power Station and TS 

[technical specification] 3.4.11 for LSCS 

[LaSalle County Station] may eventually 

require that inservice leakage and hydrostatic 

testing be conducted at RCS [reactor coolant 

system] temperatures greater than the average 

reactor coolant temperature for MODE 4 with 

the reactor shutdown. However, even with 

the required minimum reactor coolant 

temperatures less than or equal to the average 

reactor coolant temperature for MODE 4 with 

the reactor shutdown, maintaining RCS 


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temperatures within a small band during 

testing can be impractical. The proposed 

changes will not result in a significant 

change in the stored energy in the reactor 

vessel during the performance of the testing 

nor will it alter the way inservice leakage and 

hydrostatic testing is performed or 

significantly change the temperatures and 

pressures achieved to perform the testing. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes and additions result 

in increased system operability requirements 

above those that currently exist during the 

performance of inservice leakage and 

hydrostatic testing. The incremental increase 

in stored energy in the vessel during testing 

will be conservatively bounded by the 

consequences of the postulated main steam 

line break outside of primary containment 

and analyzed margins of safety are 

unchanged. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

EGC has reviewed the no significant 

hazards determination published on August 

21, 2006 (71 FR 48561) [for Technical 

Specification Task Force traveler TSTF–484]. 

The no significant hazards determination was 

made available on October 27, 2006 (71 FR 

63050) as part of the CLIIP [Consolidated 

Line Item Improvement Process] Notice of 

Availability. EGC has concluded that the 

determination presented in the notice is 

applicable to DNPS, Units 2 and 3; LSCS, 

Units 1 and 2; and QCNPS, Units 1 and 2; 

and the determination is hereby incorporated 

by reference to satisfy the requirements of 10 

CFR 50.91(a). 

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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Bradley Fewell, 

Associate General Counsel, Exelon 

Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Travis L. Tate. 

Exelon Generation Company, LLC, 

Docket No. 50–373 and 50–374, LaSalle 

County Station (LSCS), Units 1 and 2, 

LaSalle County, Illinois 

Date of Amendment Request: January 

12, 2015. A publicly-available version is 

in ADAMS under Accession No. 

ML15012A544. 

Description of amendment request: 

The proposed amendment would delete 

the limiting condition for operation 

(LCO) Note for Technical Specification 

(TS) Section 3.5.1, ‘‘ECCS [emergency 

core cooling system]—Operating.’’ The 

current Note allows the licensee to 

consider the low pressure coolant 

injection (LPCI) subsystem associated 

with the residual heat removal (RHR) 

system to be OPERABLE under 

specified conditions. 

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. 

No physical changes to the facility will 

occur as a result of this proposed 

amendment. The proposed change will not 

alter the physical design. Current TS note 

could make LSCS susceptible to potential 

water hammer in the RHR system if in the 

SDC [shutdown cooling] Mode of RHR in 

Mode 3 when swapping from the SDC to 

LPCI mode of RHR. The proposed LAR 

[license amendment request] will eliminate 

the risk for cavitation of the pump and 

voiding in the suction piping, thereby 

avoiding potential to damage the RHR 

system, including water hammer. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not alter the 

physical design, safety limits, or safety 

analysis assumptions associated with the 

operation of the plant. Accordingly, the 

change does not introduce any new accident 

initiators, nor does it reduce or adversely 

affect the capabilities of any plant structure, 

system, or component to perform their safety 

function. Deletion of the TS note is 

appropriate because current TSs could put 

the plant at risk for potential cavitation of the 

pump and voiding in the suction piping, 

resulting in potential to damage the RHR 

system, including water hammer. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change conforms to NRC 

regulatory guidance regarding the content of 

plant Technical Specifications. The proposed 

change does not alter the physical design, 

safety limits, or safety analysis assumptions 

associated with the operation of the plant. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above evaluation, EGC 

[Exelon Generation Company, LLC] 

concludes that the proposed amendment 

does not involve a significant hazards 

consideration under the standards set forth in 

10 CFR 50.92(c), and, according a finding of 

no significant hazards consideration is 

justified. 

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, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL, 60555. 

Branch Chief: Travis L. Tate. 

FirstEnergy Nuclear Operating 

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1, 

Ottawa County, Ohio 

Date of amendment request: 

December 31, 2014. A publicly-available 

version is in ADAMS under Accession 

No. ML14365A080. 

Description of amendment request: 

The proposed amendment would revise 

the frequency for the technical 

specification surveillance to verify that 

each containment spray system nozzle 

is unobstructed from a frequency of 10 

years to an event-based frequency. 

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 containment spray system and its 

spray nozzles are not accident initiators and 

therefore the proposed change does not 

involve a significant increase in the 

probability of an accident. The revised 

surveillance requirement will require eventbased frequency verification in lieu of a fixed 

frequency verification. The proposed change 

does not have a detrimental impact on the 

integrity of any plant structure, system, or 

component that may initiate an analyzed 

event. The proposed change will not alter the 

operation or otherwise increase the failure 

probability of any plant equipment that can 

initiate an analyzed accident. Because the 

system will continue to be available to 

perform its accident mitigation function, the 

consequences of accidents previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 


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Response: No. 

The proposed change will not physically 

alter the plant (no new or different type of 

equipment will be installed) or change the 

methods governing normal plant operation. 

The proposed change does not introduce new 

accident initiators or impact assumptions 

made in the safety analysis. Testing 

requirements continue to demonstrate that 

the limiting conditions for operation are met 

and the system components are functional. 

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 safety function of the CSS 

[containment spray system] is to spray water 

into the containment atmosphere in the event 

of a loss-of-coolant accident to prevent 

containment pressure from exceeding the 

design value and to remove fission products 

from the containment atmosphere. 

The CSS is not susceptible to corrosioninduced obstruction or obstruction from 

sources external to the system. Maintenance 

activities that unexpectedly introduce 

unretrievable foreign material into the system 

would require subsequent verification to 

ensure there is no nozzle blockage. The spray 

header nozzles are expected to remain 

unblocked and available in the event that a 

safety function is required. Therefore, the 

capacity of the system would remain 

unaffected. The proposed change does 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 this change. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop A–GO–15, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Travis L. Tate. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station, Unit 

1, Ottawa County, Ohio 

Date of amendment request: 

December 19, 2014. A publicly-available 

version is in ADAMS under Accession 

No. ML14353A349. 

Description of amendment request: 

The proposed amendment would revise 

the technical specifications (TS) to 

adopt performance-based Type C testing 

for the reactor containment, which 

would allow for extended test intervals 

for Type C valves up to 75 months, and 

corrects an editorial issue in the TS. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment adopts the NRGaccepted guidelines of [Nuclear Energy 

Institute] NEI 94–01, Revision 3–A, ‘‘Industry 

Guideline for Implementing PerformanceBased Option of 10 CFR part 50, Appendix 

J,’’ for [Davis-Besse Nuclear Power Station] 

DBNPS performance-based Type C 

containment isolation valve testing. Revision 

3–A of NEI 94–01 allows, based on previous 

valve leak test performance, an extension of 

Type C containment isolation valve leak test 

intervals. Since the change involves only 

performance-based Type C testing, the 

proposed amendment does not involve either 

a physical change to the plant or a change in 

the manner in which the plant is operated or 

controlled. 

Implementation of these guidelines 

continues to provide adequate assurance that 

during design basis accidents, the 

components of the primary containment 

system will limit leakage rates to less than 

the values assumed in the plant safety 

analyses. 

The proposed amendment will not change 

the leakage rate acceptance requirements. As 

such, the containment will continue to 

perform its design function as a barrier to 

fission product releases. 

Therefore, the proposed amendment does 

not significantly increase the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment to revise the 

extended frequency performance-based Type 

C testing program does not change the design 

or operation of structures, systems, or 

components of the plant. 

The proposed amendment would continue 

to ensure containment operability and would 

ensure operation within the bounds of 

existing accident analyses. There are no 

accident initiators created or affected by the 

proposed amendment. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment to revise the 

extended frequency performance-based Type 

C testing program does not affect plant 

operations, design functions, or any analysis 

that verifies the capability of a structure, 

system, or component of the plant to perform 

a design function. In addition, this change 

does not affect safety limits, limiting safety 

system setpoints, or limiting conditions for 

operation. The specific requirements and 

conditions of the Technical Specification 

Containment Leakage Rate Testing Program 

exist to ensure that the degree of containment 

structural integrity and leak-tightness that is 

considered in the plant safety analysis is 

maintained. 

The overall containment leak rate limit 

specified by Technical Specifications is 

maintained, thus ensuring the margin of 

safety in the plant safety analysis is 

maintained. The design, operation, testing 

methods, and acceptance criteria for Type A, 

Type B, and Type C containment leakage 

tests specified in applicable codes and 

standards would continue to be met with the 

acceptance of this proposed change, since 

these are not affected by this revision to the 

performance-based containment testing 

program. 

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: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop A–GO–15, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Travis L. Tate. 

Indiana Michigan Power Company 

(IandM), Docket Nos. 50–315 and 50– 

316, Donald C. Cook Nuclear Plant, 

Units 1 and 2, Berrien County, Michigan 

Date of amendment request: 

November 14, 2014, as supplemented by 

a letter dated February 12, 2015. 

Publicly-available versions are in 

ADAMS under Accession Nos. 

ML14324A209, and ML15050A247, 

respectively.) 

Description of amendment request: 

The proposed amendments would 

replace the current Donald C. Cook 

Nuclear Plant (CNP) Units 1 and 2 

technical specifications (TSs) limit on 

reactor coolant system (RCS) gross 

specific activity with a new limit on 

RCS noble gas specific activity. The 

noble gas specific activity limit would 

be based on a new DOSE EQUIVALENT 

XE–133 definition that would replace 

the current E-Bar average disintegration 

energy definition. In addition, the 

current DOSE EQUIVALENT I–131 

definition would be revised to allow the 

use of additional thyroid dose 

conversion factors. The proposed RCS 

specific activity changes are consistent 

with NRC-approved Industry Technical 


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Specification Task Force (TSTF) 

Standard Technical Specification 

change traveler, TSTF–490, Revision 0, 

‘‘Deletion of E-Bar Definition and 

Revision to Reactor Coolant System 

Specific Activity Technical 

Specification,’’ with deviations. 

Additionally, the proposed amendments 

would revise the CNP Units 1 and 2 

licensing basis and TSs to adopt the 

alternative source term (AST) as 

allowed in 10 CFR 50.67. The proposed 

amendments represent full 

implementation of the AST as described 

in the NRC’s Regulatory Guide 1.183, 

‘‘Alternative Radiological Source Terms 

for Evaluating Design Basis Accidents at 

Nuclear Power Reactors,’’ Revision 0. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration. The licensee concluded 

that the no significant hazards 

consideration determination published 

on March 19, 2007 (72 FR 12838), 

‘‘Notice of Availability of the Model 

Safety Evaluation,’’ is applicable. This 

determination is presented below, along 

with the licensee’s analysis of the 

implementation of the AST. 

Criterion 1—The Proposed Change Does 

Not Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated. 

Reactor coolant specific activity is not an 

initiator for any accident previously 

evaluated. The Completion Time when 

primary coolant gross activity is not within 

limit is not an initiator for any accident 

previously evaluated. The current variable 

limit on primary coolant iodine 

concentration is not an initiator to any 

accident previously evaluated. As a result, 

the proposed change does not significantly 

increase the probability of an accident. The 

proposed change will limit primary coolant 

noble gases to concentrations consistent with 

the accident analyses. The proposed change 

to the Completion Time has no impact on the 

consequences of any design basis accident 

since the consequences of an accident during 

the extended Completion Time are the same 

as the consequences of an accident during 

the Completion Time. As a result, the 

consequences of any accident previously 

evaluated are not significantly increased. 

There are no physical changes to the plant 

being introduced by the proposed changes to 

the accident source term. Implementation of 

AST and the associated proposed TS changes 

and new atmospheric dispersion factors have 

no impact on the probability for initiation of 

any DBAs [Design Basis Accidents]. Once the 

occurrence of an accident has been 

postulated, the new accident source term and 

atmospheric dispersion factors are an input 

to analyses that evaluate the radiological 

consequences. The proposed changes do not 

involve a revision to the design or manner in 

which the facility is operated that could 

increase the probability of an accident 

previously evaluated in Chapter 14 of the 

UFSAR. 

Based on the AST analyses, there are no 

proposed changes to performance 

requirements and no proposed revision to the 

parameters or conditions that could 

contribute to the initiation of an accident 

previously discussed in Chapter 14 of the 

UFSAR. Plant-specific radiological analyses 

have been performed using the AST 

methodology and new X/Qs have been 

established. Based on the results of these 

analyses, it has been demonstrated that the 

CR [control room] and off-site dose 

consequences of the limiting events 

considered in the analyses meet the 

regulatory guidance provided for use with 

the AST, and the doses are within the limits 

established by 10 CFR 50.67. 

Therefore, it is concluded that the 

proposed amendment does not involve a 

significant increase in the probability or the 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident from any 

Previously Evaluated. 

The proposed change in specific activity 

limits does not alter any physical part of the 

plant nor does it affect any plant operating 

parameter. The change does not create the 

potential for a new or different kind of 

accident from any previously calculated. 

No new modes of operation are introduced 

by the proposed changes. The proposed 

changes will not create any failure mode not 

bounded by previously evaluated accidents. 

Implementation of AST and the associated 

proposed TS changes and new X/Qs have no 

impact to the initiation of any DBAs. These 

changes do not affect the design function or 

modes of operation of structures, systems and 

components in the facility prior to a 

postulated accident. Since structures, 

systems and components are operated no 

differently after the AST implementation, no 

new failure modes are created by this 

proposed change. The alternative source term 

change itself does not have the capability to 

initiate accidents. 

Consequently, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

Criterion 3—The Proposed Change Does 

Not Involve a Significant Reduction in a 

Margin of Safety. 

The proposed change revises the limits on 

noble gas radioactivity in the primary 

coolant. The proposed change is consistent 

with the assumptions in the safety analyses 

and will ensure the monitored values protect 

the initial assumptions in the safety analyses. 

The AST analyses have been performed 

using approved methodologies to ensure that 

analyzed events are bounding and safety 

margin has not been reduced. Also, new X/ 

Qs, which are based on site specific 

meteorological data, were calculated in 

accordance with the guidance of RG 1.194 to 

utilize more recent data and improved 

calculational methodologies. The dose 

consequences of these limiting events are 

within the acceptance criteria presented in 

10 CFR 50.67. Thus, by meeting the 

applicable regulatory limits for AST, there is 

no significant reduction in a margin of safety. 

Therefore, because the proposed changes 

continue to result in dose consequences 

within the applicable regulatory limits, the 

proposed amendment does not involve a 

significant reduction in margin of safety. 

The NRC staff has reviewed the 

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 

amendments requested involve no 

significant hazards consideration. 

Attorney for licensee: Robert B. 

Haemer, Senior Nuclear Counsel, One 

Cook Place, Bridgman, MI 49106. 

NRC Branch Chief: David L. Pelton. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power Plant 

(CPNPP), Units 1 and 2, Somervell 

County, Texas 

Date of amendment request: January 

28, 2015. A publicly-available version is 

in ADAMS under Accession No. 

ML15036A032. 

Description of amendment request: 

The amendment would revise Technical 

Specification (TS) 5.5.16, ‘‘Containment 

Leakage Rate Testing Program,’’ for 

CPNPP, Units 1 and 2, to allow an 

increase in the 10 CFR part 50, 

appendix J, ‘‘Primary Reactor 

Containment Leakage Testing for WaterCooled Power Reactors,’’ Type A 

Integrated Leak Rate Test (ILRT) interval 

from a 10-year frequency to a maximum 

of 15 years and the extension of the 

containment isolation valves leakage 

Type C tests from its current 60-month 

frequency to 75 months in accordance 

with Nuclear Energy Institute (NEI) 94– 

01, Revision 3–A, ‘‘Industry Guidance 

for Implementing Performance Based 

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

July 2012 (ADAMS Accession No. 

ML12221A202), and conditions and 

limitations specified in NEI 94–01, 

Revision 2–A, ‘‘Industry Guidance for 

Implementing Performance Based 

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

October 2008 (ADAMS Accession No. 

ML100620847), in addition to 

limitations and conditions of NEI 94–01, 

Revision 3–A. The proposed change 

would also delete the listing of one-time 

exceptions previously granted to ILRT 

frequencies. 

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: 


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1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment to the TS 

involves the extension of the CPNPP, Units 

1 and 2 Type A containment test interval to 

15 years and the extension of the Type C test 

interval to 75 months. The current Type A 

test interval of 120 months (10 years) would 

be extended on a permanent basis to no 

longer than 15 years from the last Type A 

test. The current Type C test interval of 60 

months for selected components would be 

extended on a performance basis to no longer 

than 75 months. Extensions of up to nine 

months (total maximum interval of 84 

months for Type C tests) are permissible only 

for non-routine emergent conditions. The 

proposed extension does not involve either a 

physical change to the plant or a change in 

the manner in which the plant is operated or 

controlled. The containment is designed to 

provide an essentially leak tight barrier 

against the uncontrolled release of 

radioactivity to the environment for 

postulated accidents. The containment and 

the testing requirements invoked to 

periodically demonstrate the integrity of the 

containment exist to ensure the plant’s 

ability to mitigate the consequences of an 

accident, and do not involve the prevention 

or identification of any precursors of an 

accident. The change in dose risk for 

changing the Type A test frequency from 

three-per-ten years to once-per-fifteen-years, 

measured as an increase to the total 

integrated dose risk for all internal events 

accident sequences for CPNPP, of 1.00E–02 

person rem/yr [roentgen equivalent man per 

year] to 6.51 person-rem/yr for Unit 1 and 

6.53 person-rem/yr for Unit 2 using the EPRI 

[Energy Power Research Institute] guidance 

with the base case corrosion included. 

Therefore, this proposed extension does not 

involve a significant increase in the 

probability of an accident previously 

evaluated. 

As documented in NUREG–1493 [, 

‘‘Performance-Based Containment Leak-Test 

Program: Draft Report for Comment,’’ January 

1995 (not publicly available)], Type B and C 

tests have identified a very large percentage 

of containment leakage paths, and the 

percentage of containment leakage paths that 

are detected only by Type A testing is very 

small. The CPNPP, Units 1 and 2 Type A test 

history supports this conclusion. 

The integrity of the containment is subject 

to two types of failure mechanisms that can 

be categorized as: (1) Activity based, and; (2) 

time based. Activity based failure 

mechanisms are defined as degradation due 

to system and/or component modifications or 

maintenance. Local leak rate test 

requirements and administrative controls 

such as configuration management and 

procedural requirements for system 

restoration ensure that containment integrity 

is not degraded by plant modifications or 

maintenance activities. The design and 

construction requirements of the 

containment combined with the containment 

inspections performed in accordance with 

ASME [American Society of Mechanical 

Engineers] Section XI, the Maintenance Rule, 

and TS requirements serve to provide a high 

degree of assurance that the containment 

would not degrade in a manner that is 

detectable only by a Type A test. Based on 

the above, the proposed extensions do not 

significantly increase the consequences of an 

accident previously evaluated. 

The proposed amendment also deletes 

exceptions previously granted to allow onetime extensions of the ILRT test frequency for 

both Units 1 and 2. These exceptions were 

for activities that have already taken place so 

their deletion is solely an administrative 

action that has no effect on any component 

and no impact on how the units are operated. 

Therefore, the proposed change does not 

result in a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment to the TS 

involves the extension of the CPNPP, Unit 1 

and 2 Type A containment test interval to 15 

years and the extension of the Type C test 

interval to 75 months. The containment and 

the testing requirements to periodically 

demonstrate the integrity of the containment 

exist to ensure the plant’s ability to mitigate 

the consequences of an accident do not 

involve any accident precursors or initiators. 

The proposed change does not involve a 

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

different type of equipment will be installed) 

or a change to the manner in which the plant 

is operated or controlled. 

The proposed amendment also deletes 

exceptions previously granted to allow onetime extensions of the ILRT test frequency for 

both Units 1 and 2. These exceptions were 

for activities that would have already taken 

place by the time this amendment is 

approved; therefore, their deletion is solely 

an administrative action that does not result 

in any change in how the units are operated. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed amendment to TS 5.5.16 

involves the extension of the CPNPP, Units 

1 and 2 Type A containment test interval to 

15 years and the extension of the Type C test 

interval to 75 months for selected 

components. This amendment does not alter 

the manner in which safety limits, limiting 

safety system set points, or limiting 

conditions for operation are determined. The 

specific requirements and conditions of the 

TS Containment Leak Rate Testing Program 

exist to ensure that the degree of containment 

structural integrity and leak-tightness that is 

considered in the plant safety analysis is 

maintained. The overall containment leak 

rate limit specified by TS is maintained. 

The proposed change involves only the 

extension of the interval between Type A 

containment leak rate tests and Type C tests 

for CPNPP, Units 1 and 2. The proposed 

surveillance interval extension is bounded by 

the 15-year ILRT Interval and the 75-month 

Type C test interval currently authorized 

within NEI 94–01, Revision 3–A. Industry 

experience supports the conclusion that Type 

B and C testing detects a large percentage of 

containment leakage paths and that the 

percentage of containment leakage paths that 

are detected only by Type A testing is small. 

The containment inspections performed in 

accordance with ASME Section Xl, TS and 

the Maintenance Rule serve to provide a high 

degree of assurance that the containment 

would not degrade in a manner that is 

detectable only by Type A testing. The 

combination of these factors ensures that the 

margin of safety in the plant safety analysis 

is maintained. The design, operation, testing 

methods and acceptance criteria for Type A, 

B, and C containment leakage tests specified 

in applicable codes and standards would 

continue to be met, with the acceptance of 

this proposed change, since these are not 

affected by changes to the Type A and Type 

C test intervals. 

The proposed amendment also deletes 

exceptions previously granted to allow onetime extensions of the ILRT test frequency for 

both Units 1 and 2. These exceptions were 

for activities that would have already taken 

place by the time this amendment is 

approved; therefore, their deletion is solely 

an administrative action and does not change 

how the units are operated and maintained. 

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: Timothy P. 

Matthews, Esq., Morgan, Lewis and 

Bockius, 1111 Pennsylvania Avenue 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

South Carolina Electric and Gas 

Company Docket Nos.: 52–027 and 52– 

028, Virgil C. Summer Nuclear Station 

(VCSNS) Units 2 and 3, Fairfield 

County, South Carolina 

Date of amendment request: 

December 4, 2014. A publicly-available 

version is in ADAMs under Accession 

No. ML14339A637. 

Description of amendment request: 

The proposed change would amend 

Combined License (COL) Nos. NPF–93 

and NPF–94 for the Virgil C. Summer 

Nuclear Station (VCSNS) Units 2 and 3 

by changing the structure and layout of 

various areas of the annex building. The 

proposed amendment requires changes 

to the Updated Final Safety Analysis 

Report (UFSAR) in the form of 

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specific Design Control Document 

(DCD) Tier 2 information and involves 

changes to related plant-specific Tier 2* 

and Tier 1 information, with 

corresponding changes to the associated 

COL Appendix C information. 

Because, this proposed change 

requires a departure from Tier 1 

information in the Westinghouse 

Electric Company’s Advanced Passive 

1000 DCD, the licensee also requested 

an exemption from the requirements of 

the Generic DCD Tier 1 in accordance 

with 10 CFR 52.63(b)(1). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed additions of a new 

nonsafety-related battery, battery room and 

battery equipment room, the room height 

increase, the floor thickness changes, the 

relocation of a non-structural internal wall, 

and the associated wall, room and corridor 

changes within the annex building do not 

adversely affect the fire loading analysis 

durations of the affected fire zones and areas 

(i.e., the calculated fire durations remain less 

than their design values). Thus, the fire loads 

analysis is not adversely affected (i.e., 

analysis results remain acceptable). The safe 

shutdown fire analysis is not affected. The 

proposed changes to the structural 

configuration, including anticipated 

equipment loading, room height, and floor 

thickness are accounted for in the updated 

structural configuration model that was used 

to analyze the Annex Building for safe 

shutdown earthquake (SSE) and other design 

loads and load combinations, thus the 

structural analysis is not adversely affected. 

The structural analysis description and 

results in the UFSAR are unchanged. The 

relocated internal Annex Building wall is 

non-structural, thus this change does not 

affect the structural analyses for the Annex 

Building. The proposed changes do not 

involve any accident initiating event or 

component failure, thus the probabilities of 

the accidents previously evaluated are not 

affected. The rooms affected by the proposed 

changes do not contain or interface with 

safety-related equipment, thus the proposed 

changes would not affect any safety-related 

equipment or accident mitigating function. 

The radioactive material source terms and 

release paths used in the safety analyses are 

unchanged, thus the radiological releases in 

the accident analyses are not affected. 

With the conversion of an annex building 

room to a battery room, the building volume 

serviced by nuclear island nonradioactive 

ventilation system decreases by approximate 

five percent. This reduced volume is used in 

the post-accident main control room dose 

portion of the UFSAR LOCA radiological 

analysis. However, the volume decrease is 

not sufficient to change the calculated main 

control room dose reported in the UFSAR, 

and control room habitability is 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 additions of a new 

nonsafety-related battery, battery room and 

battery equipment room, the room height 

increase, the floor thickness changes, the 

relocation of a non-structural internal wall, 

and their associated wall, room and corridor 

changes do not change fire barrier 

performance, and the fire loading analyses 

results remain acceptable. The room height 

and floor thickness changes are consistent 

with the annex building configuration used 

in the building’s structural analysis. The 

relocated internal wall is non-structural, thus 

the structural analyses for the annex building 

are not affected. The affected rooms and 

associated equipment do not interface with 

components that contain radioactive 

material. The affected rooms do not contain 

equipment whose failure could initiate an 

accident. The proposed changes do not create 

a new 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. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed additions of a new 

nonsafety-related battery, battery room and 

battery equipment room, the room height 

increase, the floor thickness changes, the 

relocation of a non-structural internal wall, 

and their associated wall, room and corridor 

changes do not change the fire barrier 

performance of the affected fire areas. The 

affected rooms do not contain safety-related 

equipment, and the safe shutdown fire 

analysis is not affected. Because the proposed 

change does not alter compliance with the 

construction codes to which the annex 

building is designed and constructed, the 

proposed changes to the structural 

configuration, including anticipated 

equipment loading, room height, and floor 

thickness do not adversely affect the safety 

margins associated with the seismic Category 

II structural capability of the annex building. 

The floor areas and amounts of 

combustible material loads in affected fire 

zones and areas do not significantly change, 

such that their fire duration times remain 

within their two-hour design value, thus the 

safety margins associated with the fire loads 

analysis are not affected. 

No safety analysis or design basis 

acceptance limit/criterion is challenged or 

exceeded by the proposed changes, thus no 

margin of safety is reduced. 

Therefore, the proposed amendment does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Kathryn M. 

Sutton, Morgan, Lewis & Bockius LLC, 

1111 Pennsylvania Avenue NW., 

Washington, DC 20004–2514. 

NRC Branch Chief: Lawrence 

Burkhart. 

South Carolina Electric and Gas 

Company, Docket Nos.: 52–027 and 52– 

028, Virgil C. Summer Nuclear Station 

(VCSNS) Units 2 and 3, Fairfield 

County, South Carolina 

Date of amendment request: February 

10, 2015. A publicly-available version is 

in ADAMS under Accession No. 

ML15041A698. 

Description of amendment request: 

The proposed change would amend 

Combined License Nos. NPF–93 and 

NPF–94 for the Virgil C. Summer 

Nuclear Station (VCSNS) Units 2 and 3 

by revising Tier 2* information 

contained within the Human Factors 

Engineering Design Verification, Task 

Support Verification and Integrated 

System Validation plans. These 

documents are incorporated by 

reference into the VCSNS Units 2 and 3 

Updated Final Safety Analysis Report 

and will additionally require changes to 

be made to affected Tier 2 information. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment includes 

changes to Integrated System Validation 

(ISV) activities, which are performed on the 

AP1000 plant simulator to validate the 

adequacy of the AP1000 human systems 

interface design and confirm that it meets 

human factors engineering principles. The 

proposed changes involve administrative 

details related to performance of the ISV, and 

no plant hardware or equipment is affected 

whose failure could initiate an accident, or 

that interfaces with a component that could 

initiate an accident, or that contains 

radioactive material. Therefore, these 

changes have no effect on any accident 

initiator in the Updated Final Safety Analysis 

Report (UFSAR), nor do they affect the 

radioactive material releases in the UFSAR 

accident analysis. 


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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 amendment includes 

changes to ISV activities, which are 

performed on the AP1000 plant simulator to 

validate the adequacy of the AP1000 human 

system interface design and confirm that it 

meets human factors engineering principles. 

The proposed changes involve administrative 

details related to performance of the ISV, and 

no plant hardware or equipment is affected 

whose failure could initiate an accident, or 

that interfaces with a component that could 

initiate an accident, or that contains 

radioactive material. Although the ISV may 

identify a need to initiate changes to add, 

modify, or remove plant structures, systems, 

or components, these changes will not be 

made directly as part of the ISV. 

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 amendment includes 

changes to ISV activities, which are 

performed on the AP1000 plant simulator to 

validate the adequacy of the AP1000 human 

system interface design and confirm that it 

meets human factors engineering principles. 

The proposed changes involve administrative 

details related to performance of the ISV, and 

do not affect any safety-related equipment, 

design code compliance, design function, 

design analysis, safety analysis input or 

result, or design/safety margin. No safety 

analysis or design basis acceptance limit/

criterion is challenged or exceeded by the 

proposed changes, thus no margin of safety 

is reduced. 

Therefore, the proposed amendment does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Kathryn M. 

Sutton, Morgan, Lewis & Bockius LLC, 

1111 Pennsylvania Avenue NW., 

Washington, DC 20004–2514. 

NRC Branch Chief: Lawrence 

Burkhart. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–321 and 50–366, 

Hatch Nuclear Plant, Units 1 and 2, 

Appling County, Georgia 

Date of amendment request: October 

10, 2014. A publicly-available version is 

in ADAMS under Accession No. 

ML14288A226. 

Description of amendment request: 

The licensee requested 21 revisions to 

the Technical Specifications. The 

licensee states the changes were chosen 

to increase the consistency between the 

Hatch Technical Specifications, the 

Improved Standard Technical 

Specifications, and the Technical 

Specifications of other plants in the 

Southern Nuclear Operating Company 

fleet. A list of the requested revisions is 

included in Enclosure 1 of the 

application. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration for each of the 24 changes 

requested, which is presented below: 

2.1 TSTF–30–A, Revision 3, ‘‘Extend the 

Completion Time for Inoperable Isolation 

Valve to a Closed System to 72 Hours.’’ 

Specification 3.6.1.3, ‘‘Primary 

Containment Isolation Valves (PCIVs),’’ 

Action C, TS page 3.6–9, is revised to provide 

a 72 hour Completion Time for penetration 

flow paths with one inoperable PCIV with a 

closed system. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 extends the 

Completion Time to isolate an inoperable 

primary containment isolation valve (PCIV) 

from 4 hours to 72 hours when the PCIV is 

associated with a closed system. The PCIVs 

are not an initiator of any accident previously 

evaluated. The consequences of a previously 

evaluated accident during the extended 

Completion Time are the same as the 

consequences during the existing Completion 

Time. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change extends the 

Completion Time to isolate an inoperable 

primary containment isolation valve (PCIV) 

from 4 hours to 72 hours when the PCIV is 

associated with a closed system. The PCIVs 

serve to mitigate the potential for radioactive 

release from the primary containment 

following an accident. The design and 

response of the PCIVs to an accident are not 

affected by this change. The revised 

Completion Time is appropriate given the 

isolation capability of the closed system. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.2 TSTF–45–A, Revision 2, ‘‘Exempt 

Verification of CIVs that are Locked, Sealed 

or Otherwise Secured’’ 

The proposed change revises SRs 3.6.1.3.2 

and 3.6.1.3.3 in Specification 3.6.1.3, 

‘‘Primary Containment Isolation Valves 

(PCIVs),’’ to exempt manual PCIVs and blind 

flanges which are locked, sealed, or 

otherwise secured in position from position 

verification requirements. The proposed 

change also revises SR 3.6.4.2.1 in 

Specification 3.6.4.2, ‘‘Secondary 

Containment Isolation Valves (SCIVs),’’ to 

exempt manual SCIVs and blind flanges 

which are locked, sealed, or otherwise 

secured in position from position verification 

requirements. 

Signification Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 exempts manual 

primary containment isolation valves and 

blind flanges located inside and outside of 

containment, and manual secondary 

containment isolation valves and blind 

flanges, that are locked, sealed, or otherwise 

secured in position from the periodic 

verification of valve position required by 

Surveillance Requirements 3.6.1.3.2, 

3.6.1.3.3, and 3.6.4.2.1. The exempted valves 

and devices are verified to be in the correct 

position upon being locked, sealed, or 

secured. Because the valves and devices are 

in the condition assumed in the accident 

analysis, the proposed change will not affect 

the initiators or mitigation of any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 


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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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change exempts manual 

primary containment isolation valves and 

blind flanges located inside and outside of 

containment, and manual secondary 

containment isolation valves and blind 

flanges, that are locked, sealed, or otherwise 

secured in position from the periodic 

verification of valve position required by 

Surveillance Requirements 3.6.1.3.2, 

3.6.1.3.3, and 3.6.4.2.1. These valves and 

devices are administratively controlled and 

their operation is a non-routine event. The 

position of a locked, sealed or secured blind 

flange or valve is verified at the time it is 

locked, sealed or secured, and any changes 

to their position is performed under 

administrative controls. Industry experience 

has shown that these valves are generally 

found to be in the correct position. Since the 

change impacts only the frequency of 

verification for blind flange and valve 

position, the proposed change will provide a 

similar level of assurance of correct position 

as the current frequency of verification. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.3 TSTF–46–A, Revision 1, ‘‘Clarify the 

CIV Surveillance to Apply Only to Automatic 

Isolation Valves’’ 

The proposed change modifies SR 3.6.1.3.5 

in Specification 3.6.1.3, ‘‘Primary 

Containment Isolation Valves (PCIVs),’’ and 

SR 3.6.4.2.2, in Specification 3.6.4.2, 

‘‘Secondary Containment Isolation Valves 

(SCIVs),’’ including their associated Bases, to 

delete the requirement to verify the isolation 

time of ‘‘each power operated’’ containment 

isolation valve and only require verification 

of each ‘‘power operated automatic isolation 

valve.’’ 

Signification Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the 

requirements in Technical Specification 

Surveillance Requirements (SRs) 3.6.1.3.5 

and 3.6.4.2.2, and their associated Bases, to 

delete the requirement to verify the isolation 

time of ‘‘each power operated’’ PCIV and 

SCIV and only require verification of closure 

time for each ‘‘automatic power operated 

isolation valve.’’ The closure times for PCIVs 

and SCIVs that do not receive an automatic 

closure signal are not an initiator of any 

design basis accident or event, and therefore 

the proposed change does not increase the 

probability of any accident previously 

evaluated. The PCIVs and SCIVs are used to 

respond to accidents previously evaluated. 

Power operated PCIVs and SCIVs that do not 

receive an automatic closure signal are not 

assumed to close in a specified time. The 

proposed change does not change how the 

plant would mitigate an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not result in a 

change in the manner in which the PCIVs 

and SCIVs provide plant protection or 

introduce any new or different operational 

conditions. Periodic verification that the 

closure times for PCIVs and SCIVs that 

receive an automatic closure signal are 

within the limits established by the accident 

analysis will continue to be performed under 

SRs 3.6.1.3.5 and 3.6.4.2.2. The change does 

not alter assumptions made in the safety 

analysis, and is consistent with the safety 

analysis assumptions and current plant 

operating practice. There are also no design 

changes associated with the proposed 

changes, and the change does not involve a 

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

or different type of equipment will be 

installed). 

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. 

The proposed change provides clarification 

that only PCIVs and SCIVs that receive an 

automatic isolation signal are within the 

scope of SRs 3.6.1.3.5 and 3.6.4.2.2. The 

proposed change does not result in a change 

in the manner in which the PCIVs and SCIVs 

provide plant protection. Periodic 

verification that closure times for PCIVs and 

SCIVs that receive an automatic isolation 

signal are within the limits established by the 

accident analysis will continue to be 

performed. The proposed change does not 

affect the safety analysis acceptance criteria 

for any analyzed event, nor is there a change 

to any safety analysis limit. The proposed 

change does not alter the manner in which 

safety limits, limiting safety system settings 

or limiting conditions for operation are 

determined, nor is there any adverse effect on 

those plant systems necessary to assure the 

accomplishment of protection functions. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.4 TSTF–222–A, Revision 1, ‘‘Control Rod 

Scram Time Testing’’ 

Specification 3.1.4, ‘‘Control Rod Scram 

Times,’’ SRs 3.1.4.1 and 3.1.4.4, are revised 

to only require scram time testing of control 

rods that are in an affected core cell. The SR 

3.1.4.1 Frequency ‘‘Prior to exceeding 40% 

RTP after fuel movement within the reactor 

vessel,’’ is eliminated and a new Frequency 

is added to SR 3.1.4.4 which states, ‘‘Prior to 

exceeding 40% RTP after fuel movement 

within the affected core cell.’’ 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 clarifies the intent of 

Surveillance testing in Specification 3.1.4, 

‘‘Control Rod Scram Times.’’ The existing 

Specification wording requires control rod 

scram time testing of all control rods 

whenever fuel is moved within the reactor 

pressure vessel, even though the Technical 

Specification Bases state that control rod 

scram time testing is only required in the 

affected core cells. The Frequency of 

Surveillances 3.1.4.1 and 3.1.4.4 are revised 

to implement the Bases statement in the 

Specifications. The proposed change does 

not affect any plant equipment, test methods, 

or plant operation, and are not initiators of 

any analyzed accident sequence. The control 

rods will continue to perform their function 

as designed. Operation in accordance with 

the proposed Technical Specifications will 

ensure that all analyzed accidents will 

continue to be mitigated as previously 

analyzed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 


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governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change clarifies the intent of 

Surveillance testing in Specification 3.1.4, 

‘‘Control Rod Scram Times.’’ The existing 

Specification wording requires control rod 

scram time testing of all control rods 

whenever fuel is moved within the reactor 

pressure vessel, even though the Technical 

Specification Bases state that the control rod 

scam time testing is only required in the 

affected core cells. The proposed change will 

not affect the operation of plant equipment 

or the function of any equipment assumed in 

the accident analysis. Control rod scram time 

testing will be performed following any fuel 

movement that could affect the scram time. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.5 TSTF–264–A, Revision 0, ‘‘3.3.9 and 

3.3.10—Delete Flux Monitors Specific 

Overlap Requirement SRs’’ 

The proposed change revises Specification 

3.3.1.1, ‘‘RPS Instrumentation,’’ by deleting 

Surveillances 3.3.1.1.6 and 3.3.1.1.7, which 

verify the overlap between the source range 

monitor (SRM) and the intermediate range 

monitor (IRM), and between the IRM and the 

average power range monitor (APRM). 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 eliminates two 

Surveillances Requirements (SRs) (SRs 

3.3.1.1.6 and 3.3.1.1.7) which verify the 

overlap between the source range monitor 

(SRM) and intermediate range monitor (IRM) 

and between the IRM and the average power 

range monitor (APRM). The testing 

requirement is incorporated in the existing 

Channel Check Surveillance (SR 3.3.1.1.1). 

The proposed change does not affect any 

plant equipment, test methods, or plant 

operation, and are not initiators of any 

analyzed accident sequence. The SRM, IRM, 

and APRM will continue to perform their 

function as designed. Operation in 

accordance with the proposed Technical 

Specifications will ensure that all analyzed 

accidents will continue to be mitigated as 

previously analyzed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change eliminates SRs 

3.3.1.1.6 and 3.3.1.1.7 which verify the 

overlap between the SRM and IRM and 

between the IRM and the APRM. The testing 

requirement is incorporated in the existing 

Channel Check Surveillance (SR 3.3.1.1.1). 

The proposed change will not affect the 

operation of plant equipment or the function 

of any equipment assumed in the accident 

analysis. Instrument channel overlap will 

continue to be verified under the existing 

Channel Check surveillance. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.6 TSTF–269–A, Revision 2, ‘‘Allow 

Administrative Means of Position 

Verification for Locked or Sealed Valves’’ 

The proposed change modifies 

Specification 3.6.1.3, ‘‘Primary Containment 

Isolation Valves,’’ and Specification 3.6.4.2, 

‘‘Secondary Containment Isolation Valves.’’ 

The specifications require penetrations with 

an inoperable isolation valve to be isolated 

and periodically verified to be isolated. A 

Note is added to Specification 3.6.1.3, 

Actions A and C, and Specification 3.6.4.2, 

Action A, to allow isolation devices that are 

locked, sealed, or otherwise secured to be 

verified by use of administrative means. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 modifies 

Specification 3.6.1.3, ‘‘Primary Containment 

Isolation Valves,’’ and Specification 3.6.4.2, 

‘‘Secondary Containment Isolation Valves.’’ 

The specifications require penetrations with 

an inoperable isolation valve to be isolated 

and periodically verified to be isolated. A 

Note is added to Specification 3.6.1.3, 

Actions A and C, and Specification 3.6.4.2, 

Action A, to allow isolation devices that are 

locked, sealed, or otherwise secured to be 

verified by use of administrative means. The 

proposed change does not affect any plant 

equipment, test methods, or plant operation, 

and are not initiators of any analyzed 

accident sequence. The inoperable 

containment penetrations will continue to be 

isolated, and hence perform their isolation 

function. Operation in accordance with the 

proposed Technical Specifications will 

ensure that all analyzed accidents will 

continue to be mitigated as previously 

analyzed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change will not affect the 

operation of plant equipment or the function 

of any equipment assumed in the accident 

analysis. The primary and secondary 

containment isolation valves will continue to 

be operable or will be isolated as required by 

the existing specifications. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.7 TSTF–273–A, Revision 2, ‘‘Safety 

Function Determination Program 

Clarifications’’ 

The proposed Technical Specification (TS) 

changes add explanatory text to the Bases for 

limiting condition for operation (LCO) 3.0.6 

clarifying the ‘‘appropriate LCO for loss of 

function,’’ and that consideration does not 

have to be made for a loss of power in 

determining loss of function. Explanatory 

text is also added to the programmatic 

description of the Safety Function 

Determination Program (SFDP) in 

Specification 5.5.12 to provide clarification 

of these same issues. 

Signification Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 


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three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed Technical Specification (TS) 

changes add explanatory text to the 

programmatic description of the Safety 

Function Determination Program (SFDP) in 

Specification 5.5.10 to clarify in the 

requirements that consideration does not 

have to be made for a loss of power in 

determining loss of function. The Bases for 

limiting condition for operations (LCO) 3.0.6 

are revised to provide clarification of the 

‘‘appropriate LCO for loss of function,’’ and 

that consideration does not have to be made 

for a loss of power in determining loss of 

function. The changes are editorial and 

administrative in nature, and therefore do not 

increase the probability of any accident 

previously evaluated. No physical or 

operational changes are made to the plant. 

The proposed change does not change how 

the plant would mitigate an accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes are editorial and 

administrative in nature and do not result in 

a change in the manner in which the plant 

operates. The loss of function of any specific 

component will continue to be addressed in 

its specific TS LCO and plant configuration 

will be governed by the required actions of 

those LCOs. The proposed changes are 

clarifications that do not degrade the 

availability or capability of safety related 

equipment, and therefore do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. There are no design changes 

associated with the proposed changes, and 

the changes do not involve a physical 

alteration of the plant (i.e., no new or 

different type of equipment will be installed). 

The changes do not alter assumptions made 

in the safety analysis, and are consistent with 

the safety analysis assumptions and current 

plant operating practice. Due to the 

administrative nature of the changes, they 

cannot be an accident initiator. 

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. 

The proposed changes to TS 5.5.10 are 

clarifications and are editorial and 

administrative in nature. No changes are 

made the LCOs for plant equipment, the time 

required for the TS Required Actions to be 

completed, or the out of service time for the 

components involved. The proposed changes 

do not affect the safety analysis acceptance 

criteria for any analyzed event, nor is there 

a change to any safety analysis limit. The 

proposed changes do not alter the manner in 

which safety limits, limiting safety system 

settings or limiting conditions for operation 

are determined, nor is there any adverse 

effect on those plant systems necessary to 

assure the accomplishment of protection 

functions. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.8 TSTF–283–A, Revision 3, ‘‘Modify 

Section 3.8 Mode Restriction Notes’’ 

The proposed change revises several 

Specification 3.8.1, ‘‘AC Sources— 

Operating,’’ Surveillance Notes to allow full 

or partial performance of the SRs to reestablish Operability provided an assessment 

determines the safety of the plant is 

maintained or enhanced. These Surveillances 

currently have Notes prohibiting their 

performance in Modes 1 or 2, or in Modes 

1, 2, or 3. 

SR 3.8.1.6 (ISTS SR 3.8.1.8), which tests 

the transfer of Alternating (AC) sources from 

normal to alternate offsite circuits, contains 

a Note prohibiting performance in Mode 1 or 

2. The Note is modified to state that 

performance is normally prohibited in Mode 

1 or 2 but may be performed to re-establish 

Operability provided an assessment 

determines the safety of the plant is 

maintained or enhanced. 

SR 3.8.1.7 (ISTS SR 3.8.1.9), which tests 

the ability of the emergency diesel generator 

(DG) to reject a load greater than or equal to 

its associated single largest post-accident 

load, contains a Note prohibiting 

performance in Mode 1 or 2. An exception 

is provided for the swing DG. The Note is 

modified to state that performance is 

normally prohibited in Mode 1 or 2 but may 

be performed to re- establish Operability 

provided an assessment determines the safety 

of the plant is maintained or enhanced. 

SR 3.8.1.8 (ISTS SR 3.8.1.10), which tests 

emergency DG operation following a load 

rejection of greater than or equal to 2775 kW, 

contains a Note prohibiting performance in 

Mode 1 or 2. The Note is modified to state 

that performance is normally prohibited in 

Mode 1 or 2 but portions of the SR may be 

performed to re- establish Operability 

provided an assessment determines the safety 

of the plant is maintained or enhanced. 

SR 3.8.1.9 (ISTS SR 3.8.1.11), which tests 

the response to a loss of offsite power signal, 

contains a Note prohibiting performance in 

Mode 1, 2, or 3. The Note is modified to state 

that performance is normally prohibited in 

Mode 1, 2, or 3, but portions of the SR may 

be performed to re-establish Operability 

provided an assessment determines the safety 

of the plant is maintained or enhanced. 

SR 3.8.1.10 (ISTS SR 3.8.1.12), which tests 

response to an Emergency Core Cooling 

System (ECCS) initiation signal, contains a 

Note prohibiting performance in Mode 1 or 

2. The Note is modified to state that 

performance is normally prohibited in Mode 

1 or 2, but the SR may be performed to reestablish Operability provided an assessment 

determines the safety of the plant is 

maintained or enhanced. 

SR 3.8.1.11 (ISTS SR 3.8.1.13), which tests 

that each DGs automatic trips are bypassed 

on a loss of voltage signal concurrent with an 

ECCS initiation signal, contains a Note 

prohibiting performance in Mode 1, 2, or 3. 

The Note is modified to state that 

performance is normally prohibited in Mode 

1, 2, or 3, but the SR may be performed to 

re-establish Operability provided an 

assessment determines the safety of the plant 

is maintained or enhanced. 

SR 3.8.1.12 (ISTS SR 3.8.1.14), which 

performs a 24 hour loaded test run of the DG, 

contains a Note prohibiting performance in 

Mode 1 or 2. The Note is modified to state 

that performance is normally prohibited in 

Mode 1 or 2, but the SR may be performed 

to re-establish Operability provided an 

assessment determines the safety of the plant 

is maintained or enhanced. 

SR 3.8.1.14 (ISTS SR 3.8.1.16), which 

verifies transfer from DG to offsite power, 

contains a Note prohibiting performance in 

Mode 1, 2, or 3. The Note is modified to state 

that performance is normally prohibited in 

Mode 1, 2, or 3, but portions of the SR may 

be performed to re-establish Operability 

provided an assessment determines the safety 

of the plant is maintained or enhanced. 

SR 3.8.1.15 (ISTS SR 3.8.1.17), which 

verifies than a DG operating in test mode will 

return to ready-to-load condition and 

energize the emergency load from offsite 

power on receipt of an ECCS initiation signal, 

contains a Note prohibiting performance in 

Mode 1, 2, or 3. The Note is modified to state 

that performance is normally prohibited in 

Mode 1, 2, or 3, but portions of the SR may 

be performed to re-establish Operability 

provided an assessment determines the safety 

of the plant is maintained or enhanced. 

SR 3.8.1.16 (ISTS SR 3.8.1.18), which 

verifies the interval between each sequenced 

load, contains a Note prohibiting 

performance in Mode 1, 2, or 3. The Note is 

modified to state that performance is 

normally prohibited in Mode 1, 2, or 3, but 

the SR may be performed to re-establish 

Operability provided an assessment 

determines the safety of the plant is 

maintained or enhanced. 

SR 3.8.1.17 (ISTS SR 3.8.1.19), which 

verifies the response to a loss of offsite power 

signal and Engineered Safety Features (ESF) 

actuation signal, contains a Note prohibiting 

performance in Mode 1, 2, or 3. The Note is 

modified to state that performance is 

normally prohibited in Mode 1, 2, or 3, but 

portions of the SR may be performed to reestablish Operability provided an assessment 

determines the safety of the plant is 

maintained or enhanced. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 


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1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies Mode 

restriction Notes on eleven emergency diesel 

generator (DG) Surveillances to allow 

performance of the Surveillance in whole or 

in part to re-establish emergency DG 

Operability. The emergency DGs and their 

associated emergency loads are accident 

mitigating features, and are not an initiator of 

any accident previously evaluated. As a 

result the probability of any accident 

previously evaluated is not increased. The 

proposed change allows Surveillance testing 

to be performed in whole or in part to reestablish Operability of an emergency DG. 

The consequences of an accident previously 

evaluated during the period that the 

emergency DG is being tested to re-establish 

Operability are no different from the 

consequences of an accident previously 

evaluated while the emergency DG is 

inoperable. As a result, the consequences of 

any accident previously evaluated are not 

increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The purpose of Surveillances is to verify 

that equipment is capable of performing its 

assumed safety function. The proposed 

change will only allow the performance of 

the Surveillances to re-establish Operability 

and the proposed changes may not be used 

to remove an emergency DG from service. 

The proposed changes also require an 

assessment to verify that plant safety will be 

maintained or enhanced by performance of 

the Surveillance in the normally prohibited 

Modes. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.9 TSTF–284–A, Revision 3, ‘‘Add ‘Met vs. 

Perform’ to Technical Specification 1.4, 

Frequency’’ 

The change inserts a discussion paragraph 

into Specification 1.4, and two new examples 

are added to facilitate the use and application 

of SR Notes that utilize the terms ‘‘met’’ and 

‘‘perform.’’ 

Signification Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes insert a discussion 

paragraph into Specification 1.4, and several 

new examples are added to facilitate the use 

and application of Surveillance Requirement 

(SR) Notes that utilize the terms ‘‘met’’ and 

‘‘perform’’. The changes also modify SRs in 

multiple Specifications to appropriately use 

‘‘met’’ and ‘‘perform’’ exceptions. The 

changes are administrative in nature because 

they provide clarification and correction of 

existing expectations, and therefore the 

proposed change does not increase the 

probability of any accident previously 

evaluated. No physical or operational 

changes are made to the plant. The proposed 

change does not significantly change how the 

plant would mitigate an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes are administrative 

in nature and do not result in a change in the 

manner in which the plant operates. The 

proposed changes do not degrade the 

availability or capability of safety related 

equipment, and therefore do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. There are no design changes 

associated with the proposed changes, and 

the changes do not involve a physical 

alteration of the plant (i.e., no new or 

different type of equipment will be installed). 

The changes do not alter assumptions made 

in the safety analysis, and are consistent with 

the safety analysis assumptions and current 

plant operating practice. Due to the 

administrative nature of the changes, they 

cannot be an accident initiator. 

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. 

The proposed changes are administrative 

in nature and do not result in a change in the 

manner in which the plant operates. The 

proposed changes provide clarification and 

correction of existing expectations that do 

not degrade the availability or capability of 

safety related equipment, or alter their 

operation. The proposed changes do not 

affect the safety analysis acceptance criteria 

for any analyzed event, nor is there a change 

to any safety analysis limit. The proposed 

changes do not alter the manner in which 

safety limits, limiting safety system settings 

or limiting conditions for operation are 

determined, nor is there any adverse effect on 

those plant systems necessary to assure the 

accomplishment of protection functions. The 

proposed changes will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.10 TSTF–295–A, Revision 0, ‘‘Modify 

Note 2 to Actions of PAM Table to Separate 

Condition Entry for Each Penetration’’ 

Specification 3.3.3.1, ‘‘Post Accident 

Monitoring (PAM) Instrumentation,’’ 

Function 6, is renamed from ‘‘Primary 

Containment Isolation Valve Position’’ to 

‘‘Penetration Flow Path Primary Containment 

Isolation Valve Position.’’ 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 clarifies the separate 

condition entry Note in Specification 3.3.3.1, 

‘‘Post Accident Monitoring (PAM) 

Instrumentation,’’ for Function 6, ‘‘Primary 

Containment Isolation Valve Position,’’ and 

Function 9, ‘‘Suppression Pool Water 

Temperature.’’ The proposed change does not 

affect any plant equipment, test methods, or 

plant operation, and are not initiators of any 

analyzed accident sequence. The actions 

taken for inoperable PAM channels are not 

changed. Operation in accordance with the 

proposed Technical Specifications will 

ensure that all analyzed accidents will 

continue to be mitigated as previously 

analyzed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 


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governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change will not affect the 

operation of plant equipment or the function 

of any equipment assumed in the accident 

analysis. The PAM channels will continue to 

be operable or the existing, appropriate 

actions will be followed. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.11 TSTF–306–A, Revision 2, ‘‘Add Action 

to LCO 3.3.6.1 to Give Option to Isolate the 

Penetration’’ 

The proposed change revises Specification 

3.3.6.1, ‘‘Primary Containment Isolation 

Instrumentation.’’ An Actions Note is added 

allowing penetration flow paths to be 

unisolated intermittently under 

administrative controls. The traversing incore 

probe (TIP) isolation system is also 

segregated into a separate Function, allowing 

12 hours to place the channel in trip and 24 

hours to isolate the penetration. A new 

Condition G is added for the new TIP 

isolation system Function. Condition G is 

referenced from Required Action C.1 when 

Conditions A or B are not met. The 

subsequent Actions are renumbered. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.3.6.1, ‘‘Primary Containment Isolation 

Instrumentation.’’ An Actions Note is added 

allowing penetration flow paths to be 

unisolated intermittently under 

administrative controls. The traversing incore 

probe (TIP) isolation system is segregated 

into a separate Function, allowing 12 hours 

to place the channel in trip and 24 hours to 

isolate the penetration. A new Action G is 

added which is referenced by the new TIP 

isolation system Function. The subsequent 

Actions are renumbered. The proposed 

change does not affect any plant equipment, 

test methods, or plant operation, and are not 

initiators of any analyzed accident sequence. 

The allowance to unisolate a penetration 

flow path will not have a significant effect on 

mitigation of any accident previously 

evaluated because the penetration flow path 

can be isolated, if needed, by a dedicated 

operator. The option to isolate a TIP System 

penetration will ensure the penetration will 

perform as assumed in the accident analysis. 

Operation in accordance with the proposed 

Technical Specifications will ensure that all 

analyzed accidents will continue to be 

mitigated as previously analyzed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change will not affect the 

operation of plant equipment or the function 

of any equipment assumed in the accident 

analysis. The allowance to unisolate a 

penetration flow path will not have a 

significant effect on a margin of safety 

because the penetration flow path can be 

isolated manually, if needed. The option to 

isolate a TIP System penetration will ensure 

the penetration will perform as assumed in 

the accident analysis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.12 TSTF–308–A, Revision 1, 

‘‘Determination of Cumulative and Projected 

Dose Contributions in RECP’’ 

The proposed change revises Specification 

5.5.4, ‘‘Radioactive Effluent Controls 

Program,’’ paragraph e, to describe the 

original intent of the dose projections. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

5.5.4, ‘‘Radioactive Effluent Controls 

Program,’’ paragraph e, to describe the 

original intent of the dose projections. The 

cumulative and projection of doses due to 

liquid releases are not an assumption in any 

accident previously evaluated and have no 

effect on the mitigation of any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change revises Specification 

5.5.4, ‘‘Radioactive Effluent Controls 

Program,’’ paragraph e, to describe the 

original intent of the dose projections. The 

cumulative and projection of doses due to 

liquid releases are administrative tools to 

assure compliance with regulatory limits. 

The proposed change revises the requirement 

to clarify the intent, thereby improving the 

administrative control over this process. As 

a result, any effect on the margin of safety 

should be minimal. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.13 TSTF–318–A, Revision 0, ‘‘Revise 

3.5.1 for One LPCI Pump Inoperable in Each 

of Two ECCS Divisions’’ 

The proposed change adds a provision to 

Condition A of Specification 3.5.1, ‘‘ECCS— 

Operating,’’ to allow one Low Pressure 

Coolant Injection (LPCI) pump to be 

inoperable in each subsystem for a period of 

seven days. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

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 adds a provision to 

Condition A of Technical Specification (TS) 

3.5.1 to allow one Low Pressure Coolant 

Injection (LPCI) pump to be inoperable in 

each subsystem for a period of seven days. 

The change to allow one LPCI pump to be 

inoperable in both subsystems is more 

reliable than what is currently allowed by 

Condition A, which requires entry into 


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shutdown limiting condition for operation 

(LCO) 3.0.3 under these conditions. The LPCI 

mode of the Residual Heat Removal system 

is not assumed to be initiator of any analyzed 

event sequence. The consequences of an 

accident previously evaluated under the 

proposed allowance are no different than the 

consequences under the existing 

requirements. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change adds a provision to 

Condition A of Technical Specification TS 

3.5.1 to allow one LPCI pump to be 

inoperable in each subsystem for a period of 

seven days. The change to allow one LPCI 

pump to be inoperable in both subsystems is 

more reliable than what is currently allowed 

by Condition A, which requires entry into 

shutdown LCO 3.0.3 under these conditions. 

The proposed change does not affect any 

safety analysis assumptions. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.14 TSTF–322–A, Revision 2, ‘‘Secondary 

Containment and Shield Building Boundary 

Integrity SRs’ 

The proposed change revises Specification 

3.6.4.1, ‘‘Secondary Containment,’’ SRs 

3.6.4.1.3 and 3.6.4.1.4 to clarify the intent of 

the Surveillances. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.6.4.1, ‘‘Secondary Containment,’’ 

Surveillance Requirements (SRs) 3.6.4.1.3 

and 3.6.4.1.4 to clarify the intent of the 

Surveillances. The secondary containment 

and the standby gas treatment (SGT) system 

are not initiators of any accident previously 

evaluated. Operation in accordance with the 

proposed Technical Specifications will 

ensure that all analyzed accidents will 

continue to be mitigated as previously 

analyzed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change is an clarification of 

the intent of the surveillances to ensure that 

the secondary containment is not 

inappropriately declared inoperable when a 

SGT subsystem is inoperable. The safety 

functions of the secondary containment and 

the SGT system are not affected. This change 

is a correction that ensures that the intent of 

the secondary containment surveillances is 

clear. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.15 TSTF–323–A, Revision 0, ‘‘EFCV 

Completion Time to 72 hours’’ 

The proposed change revises Specification 

3.6.1.3, ‘‘Primary Containment Isolation 

Valves,’’ Action C, to provide a 72 hour 

Completion Time instead of a 12 hour 

Completion Time to isolate an inoperable 

excess flow check valve (EFCV). 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.6.1.3, ‘‘Primary Containment Isolation 

Valves,’’ Action C, to provide a 72 hour 

Completion Time instead of a 12 hour 

Completion Time to isolate an inoperable 

excess flow check valve (EFCV). The primary 

containment isolation valves (PCIVs) are not 

an initiator of any accident previously 

evaluated. The consequences of a previously 

evaluated accident during the extended 

Completion Time are the same as the 

consequences during the existing Completion 

Time. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change extends the 

Completion Time to isolate an inoperable 

primary containment penetration equipped 

with an excess flow check valve from 12 

hours to 72 hours. The PCIVs serve to 

mitigate the potential for radioactive release 

from the primary containment following an 

accident. The design and response of the 

PCIVs to an accident are not affected by this 

change. The revised Completion Time is 

appropriate given the EFCVs are on 

penetrations that have been found to have 

acceptable barrier(s) in the event that the 

single isolation valve fails. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.16 TSTF–374–A, Revision 0, ‘‘Revision to 

TS 5.5.13 and Associated TS Bases for Diesel 

Fuel Oil’’ 

The proposed change revises Specification 

5.5.9, ‘‘Diesel Fuel Oil Testing Program,’’ to 

remove references to the specific American 

Society for Testing and Materials (ASTM) 

Standard from the Administrative Controls 

Section of TS, and places them in a licenseecontrolled document. Also, alternate criteria 

are added to establish the acceptability of 

new fuel oil for use prior to and following 

the addition to storage tanks. 

Signification Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 


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Response: No. 

The proposed changes remove the 

references to specific ASTM standards from 

the Administrative Controls Section of the 

Technical Specifications (TS) and place them 

in a licensee controlled document. 

Requirements to perform testing in 

accordance with the applicable ASTM 

standards is retained in the TS as are 

requirements to perform testing of both new 

and stored diesel fuel oil. Future changes to 

the licensee controlled document will be 

evaluated pursuant to the requirements of 10 

CFR 50.59 to ensure that these changes do 

not result in more than a minimal increase 

in the probability or consequences of an 

accident previously evaluated. In addition, 

tests used to establish the acceptability of 

new fuel oil for use prior to and following 

the addition to storage tanks has been 

expanded to recognize more rigorous testing 

of water and sediment content. Relocating 

the specific ASTM standard references from 

the TS to a licensee controlled document and 

allowing a water and sediment content test 

to be performed to establish the acceptability 

of new fuel oil will not affect nor degrade the 

ability of the emergency diesel generators 

(EDGs) to perform their specified safety 

function. Fuel oil quality will continue to be 

tested and maintained to ASTM 

requirements. Diesel fuel oil testing is not an 

initiator of any accident previously 

evaluated, and the proposed changes do not 

adversely affect any accident initiators or 

precursors, or alter design assumptions, 

conditions, and configuration of the facility, 

or the manner in which the plant is operated. 

The proposed changes do not adversely affect 

the ability of structures, systems, and 

components to perform their intended safety 

function to mitigate the consequences of an 

initiating event within the assumed 

acceptance limits. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of any 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 remove the 

references to specific ASTM standards from 

the Administrative Controls Section of TS 

and place them in a licensee controlled 

document. In addition, the tests used to 

establish the acceptability of new fuel oil for 

use prior to and following the addition to 

storage tanks has been expanded to allow a 

water and sediment content test to be 

performed to establish the acceptability of 

new fuel oil. 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. The 

requirements retained in the TS will 

continue to require testing of new and stored 

diesel fuel oil to ensure the proper 

functioning of the EDGs. 

Therefore, the changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes remove the 

references to specific ASTM standards from 

the Administrative Controls Section of TS 

and place them in a licensee controlled 

document. Instituting the proposed changes 

will continue to ensure the use of applicable 

ASTM standards to evaluate the changes to 

the licensee-controlled document are 

performed in accordance with the provisions 

of 10 CFR 50.59. This approach provides an 

effective level of regulatory control and 

ensures that diesel fuel oil testing is 

conducted such that there is no significant 

reduction in a margin of safety. The margin 

of safety provided by the EDGs is unaffected 

by the proposed changes since TS 

requirements will continue to ensure fuel oil 

is of the appropriate quality. The proposed 

changes provide the flexibility needed to 

improve fuel oil sampling and analysis 

methodologies while maintaining sufficient 

controls to preserve the current margins of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.17 TSTF–400–A, Revision 1, ‘‘Clarify SR 

on Bypass of DG Automatic Trips’’ 

The proposed change revises Specification 

3.8.1, ‘‘AC Sources—Operating,’’ 

Surveillance 3.8.1.11, to clarify that the 

intent of the SR is to test the non-critical 

emergency DG automatic trips. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This change clarifies the purpose of 

Surveillance Requirement (SR) 3.8.1.11, 

which is to verify that non-critical automatic 

emergency diesel generator (DG) trips are 

bypassed in an accident. The non-critical 

automatic DG trips and their bypasses are not 

initiators of any accident previously 

evaluated. Therefore, the probability of any 

accident is not significantly increased. 

Additionally, the function of the emergency 

DG in mitigating accidents is not changed. 

The revised SR continues to ensure the 

emergency DG will operate as assumed in the 

accident analysis. 

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. 

This change clarifies the purpose of SR 

3.8.1.11, which is to verify that non-critical 

automatic emergency DG trips are bypassed 

in an accident. The proposed change does 

not involve a physical alteration of the plant 

(no new or different type of equipment will 

be installed), or a change in the methods 

governing normal plant operation. Thus, this 

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. 

This change clarifies the purpose of SR 

3.8.1.11, which is to verify that non-critical 

automatic DG trips are bypassed in an 

accident. This change clarifies the purpose of 

the SR, which is to verify that the emergency 

DG is capable of performing the assumed 

safety function. The safety function of the 

emergency DG is unaffected, so the change 

does not affect the margin of safety. 

Therefore, this change does not involve a 

significant reduction in a margin of safety. 

Based on the above, SNC concludes that 

the proposed change presents no significant 

hazards consideration under the standards 

set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.18 TSTF–439–A, Revision 2, ‘‘Eliminate 

Second Completion Times Limiting Time 

From Discovery of Failure To Meet an LCO’’ 

Specifications 3.1.7, ‘‘Standby Liquid 

Control (SLC) System;’’ 3.6.4.3, ‘‘Standby Gas 

Treatment (SGT) System;’’ 3.8.1, ‘‘AC 

Sources—Operating;’’ and 3.8.7, 

‘‘Distribution Systems—Operating,’’ contain 

Required Actions with a second Completion 

Time to establish a limit on the maximum 

time allowed for any combination of 

Conditions that result in a single continuous 

failure to meet the LCO. These Completion 

Times (henceforth referred to as ‘‘second 

Completion Times’’) are joined by an ‘‘AND’’ 

logical connector to the Condition-specific 

Completion Time and state ‘‘X days from 

discovery of failure to meet the LCO’’ (where 

‘‘X’’ varies by specification). The proposed 

change deletes these second Completion 

Times from the affected Required Actions. It 

also revises ISTS Example 1.3–3 to remove 

the discussion of second Completion Times 

and to revise the discussion in that Example 

to state that alternating between Conditions 

in such a manner that operation could 

continue indefinitely without restoring 

systems to meet the LCO is inconsistent with 

the basis of the Completion Times and is 

inappropriate. Therefore, the licensee shall 

have administrative controls to limit the 

maximum time allowed for any combination 

of Conditions that result in a single 

contiguous occurrence of failing to meet the 

LCO. 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 


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The proposed change eliminates certain 

Completion Times from the Technical 

Specifications. Completion Times are not an 

initiator to any accident previously 

evaluated. As a result, the probability of an 

accident previously evaluated is not affected. 

The consequences of an accident during the 

remaining Completion Time are no different 

than the consequences of the same accident 

during the removed Completion Times. As a 

result, the consequences of an accident 

previously evaluated are not affected by this 

change. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change to delete the second 

Completion Time does not alter the manner 

in which safety limits, limiting safety system 

settings or limiting conditions for operation 

are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

of the design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.19 TSTF–458–T, Revision 0, ‘‘Removing 

Restart of Shutdown Clock for Increasing 

Suppression Pool Temperature’’ 

The proposed change revises Specification 

3.6.2.1, ‘‘Suppression Pool Average 

Temperature,’’ Required Actions D and E, to 

eliminate redundant requirements. 

Significant Hazards Consideration SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.6.2.1, ‘‘Suppression Pool Average 

Temperature,’’ Required Actions D and E, to 

eliminate redundant requirements when 

suppression pool temperature is above the 

Limiting Conditions for Operation (LCO) 

limit. Suppression pool temperature is not an 

initiator to any accident previously 

evaluated. Suppression pool temperature 

may affect the mitigation of accidents 

previously evaluated. The proposed change 

reduces the time allowed to operate with 

suppression pool temperature above the 

limit. The consequences of an accident under 

the proposed change are no different than 

under the current requirements. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change revises Specification 

3.6.2.1, ‘‘Suppression Pool Average 

Temperature,’’ Required Actions D and E, to 

eliminate redundant requirements when 

suppression pool temperature is above the 

LCO limit. The proposed change reduces the 

time allowed to operate with suppression 

pool temperature above the limit. The 

proposed revision will not adversely affect 

the margin of safety as it corrects the Actions 

to provide appropriate compensatory 

measures when suppression pool 

temperature is greater than the limit. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.20 TSTF–464–T, Revision 0, ‘‘Clarify the 

Control Rod Block Instrumentation Required 

Action’’ 

The proposed change revises Specification 

3.3.2.1, Required Action C.2.1.2 from ‘‘Verify 

by administrative methods that startup with 

RWM inoperable has not been performed in 

the last calendar year’’ to ‘‘Verify by 

administrative methods that startup with 

RWM inoperable has not been performed in 

the last 12 months.’’ 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises a Required 

Action to limit startup with the Rod Worth 

Minimizer (RWM) inoperable from once per 

calendar year to once per 12 months. The 

RWM is used to minimize the possibility and 

consequences of a control rod drop accident. 

This change clarifies the intent of the 

limitation, but does not affect the 

requirement for the RWM to be operable. As, 

over time, the number of startups with the 

RWM inoperable will not increase, the 

probability of any accident previously 

evaluated is not significantly increased. As 

the RWM is still required to be operable, the 

consequences of an any accident previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises a Required 

Action to limit startup with the Rod Worth 

Minimizer inoperable from once per calendar 

year to once per 12 months. 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 significant change in the 

methods governing normal plant operation. 

The changes do not alter assumptions made 

in the safety analysis. The proposed changes 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed 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 change revises a Required 

Action to limit startup with the Rod Worth 

Minimizer (RWM) inoperable from once per 

calendar year to once per 12 months. This 

clarifies the intent of the Required Action. 

The number of startups with RWM 

inoperable is not increased. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed change presents no significant 

hazards consideration under the standards 

set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

2.21 ISTS Adoption #1—Revise the 5.5.7 

Introductory Paragraph To Be Consistent 

With the ISTS 

The proposed change revises the 

introductory paragraph of Specification 5.5.7, 

‘‘Ventilation Filter Testing Program (VFTP),’’ 

to be consistent with the ISTS. Specific 

requirements to perform testing after 


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structural maintenance on the HEPA filter or 

charcoal adsorber housing or following 

painting, fire or chemical release, and after 

every 720 hours of operation are relocated to 

the licensee- controlled program. 

The existing wording states, ‘‘The VFTP 

will establish the required testing of 

Engineered Safety Feature (ESF) filter 

ventilation systems at the frequencies 

specified in Regulatory Guide 1.52, Revision 

2, Sections C.5.c and C.5.d, or: (1) After any 

structural maintenance on the HEPA filter or 

charcoal adsorber housings, (2) following 

painting, fire or chemical release in any 

ventilation zone communicating with the 

system, or 3) after every 720 hours of 

charcoal adsorber operation.’’ 

The proposed wording states, ‘‘A program 

shall be established to implement the 

following required testing of Engineered 

Safety Feature (ESF) filter ventilation systems 

at the frequencies specified in Regulatory 

Guide 1.52, Revision 2, Sections C.5.c and 

C.5.d, and in accordance with Regulatory 

Guide 1.52, Revision 2.’’ 

Significant Hazards Consideration: SNC 

has evaluated whether or not a significant 

hazards consideration is involved with the 

proposed amendment(s) by focusing on the 

three standards set forth in 10 CFR 50.92, 

‘‘Issuance of amendment,’’ as discussed 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the 

introductory paragraph of Specification 5.5.7, 

‘‘Ventilation Filter Testing Program (VFTP),’’ 

to be consistent with the ISTS. Specific 

requirements to perform testing after 

structural maintenance on the HEPA filter or 

charcoal adsorber housing or following 

painting, fire or chemical release, and after 

every 720 hours of operation are retained as 

a reference to Regulatory Guide requirements 

and general requirements in Surveillance 

Requirement (SR) 3.0.1. Implementation of 

these requirements will be in the licenseecontrolled VFTP. The VFTP will be 

maintained in accordance with 10 CFR 50.59. 

Since any changes to the VFTP will be 

evaluated under 10 CFR 50.59, no significant 

increase in the probability or consequences 

of an accident previously evaluated will be 

allowed. 

Therefore, this proposed change does not 

represent 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 involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

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. 

The proposed change revises the 

introductory paragraph of Specification 5.5.7, 

‘‘Ventilation Filter Testing Program (VFTP),’’ 

to be consistent with the ISTS. The proposed 

change will not reduce a margin of safety 

because it has no effect on any safety analysis 

assumption. In addition, no requirements are 

being removed, but are being replaced with 

references to an NRC Regulatory Guide and 

the requirements of SR 3.0.1. 

Therefore, this proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jennifer M. 

Buettner, Associate General Counsel, 

Southern Nuclear Operating Company, 

40 Inverness Center Parkway, 

Birmingham, AL 35201 

NRC Branch Chief: Robert J. 

Pascarelli. 

Tennessee Valley Authority, Docket 

Nos. 50–259, 50–260, and 50–296, 

Browns Ferry Nuclear Plant, Units 1, 2, 

and 3, Limestone County, Alabama 

Date of amendment request: 

December 11, 2014 (ADAMS Accession 

No. ML14349A694). 

Description of amendment request: 

The amendment would revise Section 

3.8.3, ‘‘Diesel Fuel Oil, Lube Oil, and 

Starting Air,’’ of the Technical 

Specifications (TSs) by replacing the 

current volume requirements with the 

number of continuous days the diesel 

generators (DGs) are required to run. 

The numerical volumes will be 

maintained in the licensee-controlled 

TSs Bases document so they may be 

modified under licensee control. The 

resulting requirements will specify an 

inventory of stored diesel fuel oil and 

lube oil sufficient for a 7-day supply for 

each DG. This proposed amendment is 

consistent with NRC’s approved 

Technical Specifications Task Force 

(TSTF) Improved Standard Technical 

Specifications Change Traveler TSTF– 

501, Revision 1, ‘‘Relocate Stored Fuel 

Oil and Lube Oil Volume Values to 

Licensee Control.’’ The availability of 

this TSs improvement was announced 

in the Federal Register on May 26, 2010 

(75 FR 29588). The licensee also 

proposed additional changes to Section 

3.8.3 and Section 5.5.9, ‘‘Diesel Fuel Oil 

Testing Program,’’ to support other 

related changes proposed by TSTF–501, 

Revision 1. These additional changes 

concern fuel oil quality and associated 

surveillance requirements (SRs). 

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 to TS Section 3.8.3, 

Conditions A and B, and to SR 3.8.3.1 and 

SR 3.8.3.2 remove the volume of diesel fuel 

oil and lube oil required to support 7-day 

operation of each onsite diesel generator, and 

the volume equivalent to a 6-day supply, 

from the TS and replace them with the 

associated number of days. The numerical 

volumes will be maintained under licensee 

control. The specific volume of fuel oil 

equivalent to a 7 and 6-day supply is 

calculated using the NRC-approved 

methodology described in Regulatory Guide 

1.137, Revision 1, ‘‘Fuel-Oil Systems for 

Standby Diesel Generators’’ and ANSI 

[American National Standards Institute]- 

N195 1976, ‘‘Fuel Oil Systems for Standby 

Diesel-Generators.’’ The specific volume of 

lube oil equivalent to a 7-day and 6-day 

supply is based on the diesel generator 

manufacturer’s consumption values for the 

run time of the diesel generator. Because the 

requirement to maintain a 7-day supply of 

diesel fuel oil and lube oil is not changed and 

is consistent with the assumptions in the 

accident analyses, and the actions taken 

when the volume of fuel oil and lube oil are 

less than a 6-day supply have not changed, 

neither the probability nor the consequences 

of any accident previously evaluated will be 

affected. 

The addition of a new Condition D 

provides a required action and completion 

time if new fuel oil properties are not within 

limits. The new SR 3.8.3.5 requires checking 

for and removing water from the 7-day 

storage tank every 31 days. The revised 

Section 5.5.9 adds testing requirements for 

new fuel oil to be completed prior to the 

addition of the new fuel oil to the 7-day 

storage tank, as well as additional testing to 

be completed prior or within 31 days of the 

addition. These requirements are more 

restrictive testing requirements and provide 

corrective action to be taken if the testing 

limits are not met. They are taken from the 

current NRC approved NUREG–1433, 

Revision 4, ‘‘Standard Technical 

Specifications, General Electric BWR/4 

Plants.’’ Improved, more restrictive testing 

standards will neither change the probability 

or the consequences of any accident 

previously evaluated be affected. 

Therefore, the proposed changes do not 

involve a significant increase in the 


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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 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. The change does not alter 

assumptions made in the safety analysis but 

ensures that the diesel generator operates as 

assumed in the accident analysis. The 

proposed change is consistent with the safety 

analysis assumptions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes to Section 3.8.3, 

Conditions A and B, and to SR 3.8.3.1 and 

SR 3.8.3.2 remove the numerical volume of 

diesel fuel oil and lube oil required to 

support 7-day operation of each onsite diesel 

generator, and the numerical volume 

equivalent to a 6-day supply from the TS and 

replaces them with the associated number of 

days. The numerical volumes will be 

maintained under licensee control. As the 

bases for the existing limits on diesel fuel oil 

volume and lube oil volume are not changed, 

no change is made to the accident analysis 

assumptions and no margin of safety is 

reduced as part of this change. 

The new, more restrictive, testing 

requirements, and the provision for 

corrective action to be taken if the testing 

limits are not met, are taken from the current 

NRC approved NUREG–1433, Revision 4, 

‘‘Standard Technical Specifications, General 

Electric BWR/4 Plants.’’ These changes do 

not revise the accident analysis assumptions 

and no margin of safety is reduced as part of 

these changes. 

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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, 6A West 

Tower, Knoxville, TN 37902. 

NRC Branch Chief: Shana R. Helton. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

November 20, 2014. A publiclyavailable version is in ADAMS under 

Accession No. ML14330A247. 

Description of amendment request: 

The amendment would revise the 

Technical Specification (TS) 

requirements to address NRC Generic 

Letter 2008–01, ‘‘Managing Gas 

Accumulation in Emergency Core 

Cooling, Decay Heat Removal, and 

Containment Spray Systems,’’ as 

described in Technical Specification 

Task Force (TSTF) Traveler TSTF–523, 

Revision 2, ‘‘Generic Letter 2008–01, 

Managing Gas Accumulation.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises or adds SRs 

[surveillance requirements] that require 

verification that the Emergency Core Cooling 

System (ECCS), the Residual Heat Removal 

(RHR) System, and the Containment Spray 

System are not rendered inoperable due to 

accumulated gas and to provide allowances 

which permit performance of the revised 

verification. Gas accumulation in the subject 

systems is not an initiator of any accident 

previously evaluated. As a result, the 

probability of any accident previously 

evaluated is not significantly increased. The 

proposed SRs ensure that the subject systems 

continue to be capable to perform their 

assumed safety function and are not rendered 

inoperable due to gas accumulation. Thus, 

the consequences of any accident previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the ECCS, the 

RHR System, and the Containment Spray 

System are not rendered inoperable due to 

accumulated gas and to provide allowances 

which permit performance of the revised 

verification. The proposed change does not 

involve a physical alteration of the plant (i.e., 

no new or different type of equipment will 

be installed) or a change in the methods 

governing normal plant operation. In 

addition, the proposed change does not 

impose any new or different requirements 

that could initiate an accident. The proposed 

change does not alter assumptions made in 

the safety analysis and is consistent with the 

safety analysis assumptions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the ECCS, the 

RHR System, and the Containment Spray 

System are not rendered inoperable due to 

accumulated gas and to provide allowances 

which permit performance of the revised 

verification. The proposed change adds new 

requirements to manage gas accumulation in 

order to ensure the subject systems are 

capable of performing their assumed safety 

functions. The proposed SRs are more 

comprehensive than the current SRs and will 

ensure that the assumptions of the safety 

analysis are protected. The proposed change 

does not adversely affect any current plant 

safety margins or the reliability of the 

equipment assumed in the safety analysis. 

Therefore, there are no changes being made 

to any safety analysis assumptions, safety 

limits or limiting safety system settings that 

would adversely affect plant safety as a result 

of the proposed change. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

III. Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses and Combined Licenses, 

Proposed No Significant Hazards 

Consideration Determination, and 

Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 


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Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request: August 

22, 2014. A publicly-available version is 

in ADAMS under Accession No. 

ML14237A729. 

Brief description of amendment 

request: The proposed amendment 

would revise the technical specification 

(TS) surveillance requirement (SR) for 

the ultimate heat sink (UHS) to clarify 

that spray pond level is the average of 

the level in both ponds. The design of 

the ultimate heat sink is such that it is 

difficult to meet the current SR when 

only one standby service water (SW) 

pump is in operation without 

overflowing a spray pond resulting in a 

net loss of water inventory, which may 

challenge the ability of the UHS to 

provide sufficient inventory for 30 days. 

However, if the SR is not met, a plant 

shutdown is required. 

Date of publication of individual 

notice in Federal Register: September 

5, 2014 (79 FR 53085). 

Expiration date of individual notice: 

October 6, 2014 (public comments); 

November 4, 2014 (hearing requests). 

IV. Notice of Issuance of Amendments 

to Facility Operating Licenses and 

Combined Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

A notice of consideration of issuance 

of amendment to facility operating 

license or combined license, as 

applicable, proposed no significant 

hazards consideration determination, 

and opportunity for a hearing in 

connection with these actions, was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items can be accessed as described in 

the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

DTE Electric Company, Docket No. 50– 

341, Fermi 2, Monroe County, Michigan 

Date of amendment request: April 23, 

2013, as supplemented by letters dated 

June 19, and October 13, 2014. 

Brief description of amendment: The 

amendment revised the Fermi 2 

technical specification (TS) surveillance 

requirements (SRs) associated with SR 

3.8.4.2 and SR 3.8.4.5 to add a battery 

resistance limit; SR 3.8.6.3 to change the 

average electrolyte temperature of 

representative cells, and SR 3.8.4.8 to 

change the frequency of battery capacity 

testing. 

Date of issuance: March 16, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 199. A publiclyavailable version is in ADAMS under 

Accession No. ML15057A297; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. NPF– 

43: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 22, 2014 (79 FR 42542). 

The supplemental letters dated June 19, 

and October 13, 2014, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 16, 2015. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: June 13, 

2013, as supplemented by letters dated 

August 28 and November 3, 2014, and 

January 22, 2015. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications to risk-inform 

requirements regarding selected 

Required Action end states by adopting 

Technical Specification Task Force 

(TSTF)–423, Revision 1, ‘‘Technical 

Specifications End States, NEDC– 

32998–A,’’ with some deviations as 

approved by the NRC staff. This 

technical specification improvement is 

part of the Consolidated Line Item 

Improvement Process (CLIIP). In 

addition, it approves a change to the 

facility operating license for the River 

Bend Station, Unit 1. The change 

deletes two license conditions that are 

no longer applicable and adds a new 

license condition for maintaining 

commitments required for the approval 

of this TSTF into the Updated Safety 

Analysis Report. 

Date of issuance: February 17, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 90 

days from the date of issuance. 

Amendment No.: 185. A publiclyavailable version is in ADAMS under 

Accession No. ML14106A167; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. NPF– 

47: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 20, 2013 (78 FR 

51226). The supplemental letters dated 

August 28, and November 3, 2014, and 

January 22, 2015, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 17, 

2015. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit 3, Westchester 

County, New York 

Date of amendment request: February 

4, 2014, as supplemented by letter dated 

December 9, 2014. 

Brief description of amendment: The 

amendment revised Technical 

Specification 5.5.15, ‘‘Containment 

Leakage Rate Testing Program,’’ to allow 

a permanent extension of the Type A 

primary containment integrated leak 


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rate test frequency from once every 10 

years to once every 15 years. 

Date of issuance: March 13, 2015. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 256. A publiclyavailable version is in ADAMS under 

Accession No. ML15028A308; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. DPR– 

64: The amendment revised the Facility 

Operating License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 8, 2014 (79 FR 38587). 

The supplemental letter 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 March 13, 2015. 

No significant hazards consideration 

comments received: No 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit 3, Westchester 

County, New York 

Date of amendment request: April 1, 

2014. 

Brief description of amendment: The 

amendment revised Technical 

Specification Figures 3.4.3–1, ‘‘Heatup 

Limitations for Reactor Coolant 

System,’’ 3.4.3–2, ‘‘Cooldown 

Limitations for Reactor Coolant 

System,’’ and 3.4.3–3, ‘‘Hydrostatic and 

Inservice Leak Testing Limitations for 

Reactor Coolant System’’ to address 

vacuum fill operations in the TSs. The 

proposed changes clarify that the figures 

are applicable for vacuum fill 

conditions where pressure limits are 

considered to be met for pressures that 

are below 0 pounds per square inch 

gauge (psig) (i.e., up to and including 

full vacuum conditions). Vacuum fill 

operations for the RCS can result in 

system pressures below 0 psig. 

Date of issuance: March 6, 2015. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 255. A publiclyavailable version is in ADAMS under 

Accession No. ML15050A144; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. DPR– 

64: The amendment revised the Facility 

Operating License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 28, 2014 (79 FR 

64223). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 6, 2015. 

No significant hazards consideration 

comments received: No 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request: April 5, 

2013, as supplemented by letter dated 

March 20, 2014. 

Brief description of amendment: This 

amendment revised Technical 

Specification (TS) 2.1.1 and 2.1.2, 

‘‘Safety Limits,’’ by reducing the reactor 

steam dome pressure from 785 pounds 

per square inch gauge (psig) to 685 psig 

to resolve the Pressure Regulator 

Failure-Open transient. 

Date of issuance: March 12, 2015. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days of issuance. 

Amendment No.: 242. A publiclyavailable version is in ADAMS under 

Accession No. ML14272A070; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–35: Amendment revised the 

License and TS. 

Date of initial notice in Federal 

Register: August 6, 2013 (78 FR 47788). 

The supplement dated March 20, 2014, 

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 March 12, 2015. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station, Unit 1, Oswego County, 

New York 

Date of application for amendment: 

March 8, 2013, as supplemented by 

letter dated May 16, 2013, July 8, July16, 

August 29, 2014, and January 22, 2015. 

The public versions of these documents 

are available in ADAMS at the 

Accession Nos. ML13073A103, 

ML13144A068, ML14203A050, 

ML14199A384, ML14251A233, and 

ML15026A132, respectively. 

Brief description of amendment: The 

amendment to the Nine Mile Point Unit 

1 (NMP1) Renewed Facility Operating 

License DPR–63 modified Technical 

Specification (TS) Table 3.6.2i, ‘‘Diesel 

Generator Initiation,’’ by revising the 

existing 4.16kV Power Board (PB) 102/ 

103 Emergency Bus Undervoltage 

(Degraded Voltage) Operating Time 

value and by updating the Set Point 

heading title. The TS revisions are being 

made to resolve the green non-cited 

violation (NCV) associated with the vital 

bus degraded voltage protection time 

delay documented in NRC Inspection 

Report (IR) 05000220/201101, ‘‘Nine 

Mile Point Nuclear Station—NRC 

Unresolved Item Follow-up Inspection 

Report,’’ dated January 23, 2012 

(ADAMS Accession No. ML12023A119), 

specifically, NCV05000220/20 11011– 

01, ‘‘Vital Bus Degraded Voltage Time 

Delay Not Maintained within LOCA 

Analysis Assumptions.’’ 

Date of issuance: March 12, 2015. 

Effective date: effective as of the date 

of its issuance and shall be 

implemented within 60 days. 

Amendment No.: 217. 

Renewed Facility Operating License 

No. DPR–63: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: June 11, 2013, (78 FR 35062). 

The supplements dated May 16, 2013, 

July 8, July16, August 29, 2014, and 

January 22, 2015, 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 

initial proposed no significant hazards 

consideration determination noticed in 

the Federal Register on June 11, 2013 

(78 FR 35062). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 12, 2015. 

No significant hazards consideration 

comments received: No 

Exelon Generation Company, LLC, and 

PSEG Nuclear LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania 

Date of application for amendments: 

July 11, 2014, as supplemented by letter 

dated December 1, 2014. 

Brief description of amendments: The 

amendments incorporate several 

administrative changes to the Facility 

Operating Licenses (FOLs) and the 

Technical Specifications (TSs) such as 

deleting historical items that are no 

longer applicable, correcting errors, and 

removing references that are no longer 

valid. 

Date of issuance: March 11, 2015. 


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Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendments Nos.: 296 and 299. A 

publicly-available version is in ADAMS 

under Accession No. ML14363A227; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Renewed Facility Operating License 

Nos. DPR–44 and DPR–56: The 

amendments revised the FOLs and the 

TSs. 

Date of initial notice in Federal 

Register: September 2, 2014 (79 FR 

52062). The supplemental letter dated 

December 1, 2014, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 11, 2015. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Units 1 and 2 (BVPS–1 and 2), 

Beaver County, Pennsylvania 

Date of amendment request: October 

18, 2013, as supplemented by letters 

dated June 26, 2014, September 21, 

2014, and February 4, 2015. 

Brief description of amendments: The 

amendment changes the Beaver Valley 

Power Station Technical Specifications 

(TS). Specifically, this change request 

involves the adoption of an approved 

change to the standard TS for 

Westinghouse plants (NUREG–1431), to 

allow relocation of specific TS 

surveillance frequencies to a licenseecontrolled program. The proposed 

change is described in TS Task Force 

(TSTF) Traveler, TSTF–425, Revision 3, 

‘‘Relocation Surveillance Frequencies to 

Licensee Control—RITSTF [RiskInformed Technical Specifications Task 

Force] Initiative 5b’’ (Agencywide 

Documents Access and Management 

System (ADAMS) Accession No. 

ML090850642). A Notice of Availability 

was published in the Federal Register 

on July 6, 2009 (74 FR 31996). 

The proposed change relocates 

surveillance frequencies to a licenseecontrolled program, the Surveillance 

Frequency Control Program. This 

change is applicable to licensees using 

probabilistic risk guidelines contained 

in NRC-approved NEI 04–10, Revision 

1, ‘‘Risk-Informed Technical 

Specifications Initiative 5b, RiskInformed Method for Control of 

Surveillance Frequencies’’ (ADAMS 

Accession No. ML071360456). 

Date of issuance: March 6, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days of issuance. 

Amendment Nos.: 292 and 179. A 

publicly-available version is in ADAMS 

under Accession No. ML14322A461; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Renewed Facility Operating License 

Nos. DPR–66 and NPF–73: Amendments 

revised the Renewed Facility Operating 

Licenses and Technical Specifications. 

Date of initial notice in Federal 

Register: January 21, 2014 (79 FR 

3416). The supplemental letters dated 

June 26, 2014, September 21, 2014, and 

February 4, 2015, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 6, 2015. 

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: 

November 21, 2013, and supplemented 

by the letters dated March 5 and June 

30, 2014. 

Brief description of amendment: The 

amendment authorizes changes to the 

VEGP Units 3 and 4 Updated Final 

Safety Analysis Report to revise the 

details of the effective thermal 

conductivity resulting from the 

oxidation of the inorganic zinc 

component of the containment vessel 

coating system. 

Date of issuance: February 26, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 31. A publiclyavailable version is in ADAMS under 

Accession No. ML15028A358; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Combined Licenses Nos. NPF– 

91 and NPF–92: Amendment revised the 

Facility Combined Licenses. 

Date of initial notice in Federal 

Register: March 18, 2014 (79 FR 

15150). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 26, 

2015. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of application for amendment: 

September 25, 2012; as supplemented 

on December 20, 2012; September 16, 

October 30, and November 12, 2013; 

April 23, May 23, July 3, August 11, 

August 29, and October 13, 2014; and 

January 16, 2015. 

Brief description of amendments: The 

amendment authorizes the transition of 

the Joseph M. Farley Nuclear Plant, 

Units 1 and 2, fire protection program 

to a risk-informed, performance-based 

program based on National Fire 

Protection Association (NFPA) 805, 

‘‘Performance-Based Standard for Fire 

Protection for Light Water Reactor 

Electric Generating Plants, 2001 

Edition’’ (NFPA 805), in accordance 

with 10 CFR 50.48(c). 

Date of issuance: March 10, 2015. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: Unit 1–196, Unit 

2–192. A publicly-available version is in 

ADAMS under Accession No. 

ML14308A048, documents related to 

these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Facility Operating License Nos. NPF– 

2 and NPF–8: The amendments revised 

the Renewed Facility Operating 

Licenses and Technical Specifications. 

Date of initial notice in Federal 

Register: March 12, 2013 (78 FR 

15750). The supplemental letters dated 

September 16, October 30, and 

November 12, 2013; April 23, May 23, 

July 3, August 11, August 29, and 

October 13, 2014; and January 16, 2015, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 10, 2015. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 23rd day 

of March 2015. 


asabaliauskas on DSK5VPTVN1PROD with NOTICES
Federal Register / Vol. 80, No. 61 / Tuesday, March 31, 2015 / Notices 17109 

For the Nuclear Regulatory Commission. 

Michele G. Evans, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2015–07192 Filed 3–30–15; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2015–0001] 

Sunshine Act Meeting Notice 

DATE: March 30, April 6, 13, 20, 27, 

May 4, 2015. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

Week of March 30, 2015 

There are no meetings scheduled for 

the week of March 30, 2015. 

Week of April 6, 2015—Tentative 

There are no meetings scheduled for 

the week of April 6, 2015. 

Week of April 13, 2015—Tentative 

Tuesday, April 14, 2015 

9:30 a.m. Meeting with the Advisory 

Committee on the Medical Uses of 

Isotopes 

(Public Meeting) 

(Contact: Nima Ashkeboussi, 301- 

415–5775) 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov/. 

Thursday, April 16, 2015 

9:30 a.m. Meeting with the Organization 

of Agreement States and the 

Conference of Radiation Control 

Program Directors 

(Public Meeting) 

(Contact: Nima Ashkeboussi, 301– 

415–5775) 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov/. 

Week of April 20, 2015—Tentative 

There are no meetings scheduled for 

the week of April 20, 2015. 

Week of April 27, 2015—Tentative 

There are no meetings scheduled for 

the week of April 27, 2015. 

Week of May 4, 2015—Tentative 

There are no meetings scheduled for 

the week of May 4, 2015. 

* * * * * 

The schedule for Commission 

meetings is subject to change on short 

notice. For more information or to verify 

the status of meetings, contact Glenn 

Ellmers at 301–415–0442 or via email at 

Glenn.Ellmers@nrc.gov. 

* * * * * 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: http://www.nrc.gov/public-involve/

public-meetings/schedule.html. 

* * * * * 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

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 301–287–0727, 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 

Brenda.Akstulewicz@nrc.gov or 

Patricia.Jimenez@nrc.gov. 

Dated: March 26, 2015. 

Glenn Ellmers, 

Policy Coordinator, Office of the Secretary. 

[FR Doc. 2015–07384 Filed 3–27–15; 11:15 am] 

BILLING CODE 7590–01–P 

OFFICE OF PERSONNEL 

MANAGEMENT 

Civilian Acquisition Workforce 

Personnel Demonstration Project; 

Department of Defense 

AGENCY: U.S. Office of Personnel 

Management (OPM). 

ACTION: Notice of amendments to the 

project plan for the Department of 

Defense (DoD) Civilian Acquisition 

Workforce Personnel Demonstration 

Project (AcqDemo). 

SUMMARY: The DoD, with the approval of 

OPM, received authority to conduct a 

personnel demonstration project within 

DoD’s civilian acquisition workforce 

and among those supporting personnel 

assigned to work directly with it. This 

notice announces the repeal and 

replacement of AcqDemo’s original legal 

authorization and modifies the project 

plan to include new provisions; updates 

the project plan to address changes 

resulting from new General Schedule 

regulations and operational experience; 

announces guidelines for a formal 

process for interested DoD civilian 

acquisition organizations to use to 

request approval to participate in 

AcqDemo; and provides notice of 

expansion of coverage to new or 

realigned organizations. 

DATES: The amendments will become 

effective as of March 31, 2015. 

FOR FURTHER INFORMATION CONTACT: (1) 

DoD: Darryl R. Burgan, Civilian 

Acquisition Workforce Personnel 

Demonstration Project Program Office, 

9820 Belvoir Road, Ft. Belvoir, VA 

22060, (703) 805–5050; (2) OPM: Zelma 

Moore, U.S. Office of Personnel 

Management, 1900 E Street NW., Room 

7456, Washington, DC 20415, (202) 606– 

1157. 

SUPPLEMENTARY INFORMATION: 

A. Background 

The AcqDemo Project was established 

under the authority of the Secretary of 

Defense, with the approval of OPM. 

Subject to the authority, direction, and 

control of the Secretary, the Under 

Secretary of Defense for Acquisition, 

Technology, and Logistics (USD(AT&L)) 

carries out the powers, functions, and 

duties of the Secretary concerning the 

DoD acquisition workforce. As stated in 

the most recent legislative 

authorization, the purpose of the 

demonstration project is ‘‘to determine 

the feasibility or desirability of one or 

more proposals for improving the 

personnel management policies or 

procedures that apply with respect to 

the acquisition workforce of the [DoD] 

and supporting personnel assigned to 

work directly with the acquisition 

workforce.’’ 

This demonstration project was 

originally authorized under section 

4308 of the National Defense 

Authorization Act (NDAA) for Fiscal 

Year (FY) 1996 (Pub. L. 104–106, 110 

Stat. 669; 10 United States Code 

Annotated (U.S.C.A.) 1701 note), as 

amended by section 845 of NDAA for 

FY 1998 (Pub. L. 105–85, 111 Stat.1845); 

section 813 of NDAA for FY 2003 (Pub. 

L. 107–314, 116 Stat. 2609); and section 

1112 of NDAA for FY 2004 (Pub. L. 

108–136, 117 Stat. 1634). Section 1113 

of NDAA for FY 2010 (Pub. L. 111–84, 

123 Stat. 2190) repealed the National 

Security Personnel System and directed 

conversion of all NSPS employees to 

their previous pay system by January 1, 

2012. All NSPS employees formerly in 

AcqDemo were transitioned back to 

AcqDemo during the month of May 

2011. On January 7, 2011, the original 

demonstration project authority was 

repealed and codified at section 1762 of 


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