Federal Register / Vol. 75, No. 182 / Tuesday, September 21, 2010 / Notices 57521 

NATIONAL SCIENCE FOUNDATION 

Networking and Information 

Technology Research and 

Development (NITRD) Program: Draft 

NITRD 2010 Strategic Plan—URL 

Correction 

AGENCY: The National Coordination 

Office (NCO) for Networking and 

Information Technology Research and 

Development (NITRD). 

ACTION: Notice, request for public 

comment. 

FOR FURTHER INFORMATION CONTACT: The 

National Coordination Office (NCO) at 

nitrd-sp@nitrd.gov or (703) 292–4873. 

Individuals who use a 

telecommunications device for the deaf 

(TDD) may call the Federal Information 

Relay Service (FIRS) at 1–800–877–8339 

between 8 a.m. and 8 p.m., Eastern time, 

Monday through Friday. 

DATES: Comments must be received by 

5 p.m. EDT on October 11, 2010. 

SUMMARY: With this notice, the National 

Coordination Office for Networking and 

Information Technology Research and 

Development (NITRD) requests 

comments from the public regarding the 

draft 2010 Strategic Plan for the Federal 

NITRD Program. The draft Strategic Plan 

is posted at: http://www.nitrd.gov/ 

DraftStrategicPlan/. Comments of one 

page or less in length are requested. 

This request for information will be 

active from September 10, 2010 to 

October 11, 2010. 

ADDRESSES: Submit comments via email to: nitrd-sp@nitrd.gov. Comments 

submitted in response to this notice may 

be made available to the public online 

or by alternative means. For this reason, 

please do not include in your comments 

information of a confidential nature, 

such as sensitive personal information 

or proprietary information. 

SUPPLEMENTARY INFORMATION: 

Overview: This notice is issued by the 

National Coordination Office for the 

Networking and Information 

Technology Research and Development 

(NITRD) Program. The draft NITRD 

Strategic Plan reflects broad input from 

Federal agencies as well as from 

researchers and other stakeholders in 

academia, industry, national 

laboratories, and professional/technical 

organizations. Public inputs were 

solicited in a detailed August 2008 

Request for Information (RFI) and in a 

February 2009 public forum and 

Webcast. Several hundred comments 

were received in response to the RFI, 

and many of these were posted to the 

NITRD Web site for further comment. 

The public forum, which included 

formal presentations by academic and 

industry experts addressing key 

concepts for the draft Strategic Plan, 

was attended by some 100 members of 

the public, while another 400 persons 

participated via the Webcast. 

Background: As required by the HighPerformance Computing Act of 1991 

(Pub. L. 102–194), the Next Generation 

Internet Research Act of 1998 (Pub. L. 

105–305), and the America COMPETES 

Act of 2007 (Pub. L. 110–69), NITRD 

currently provides a framework and 

mechanisms for coordination among 14 

Federal agencies that support advanced 

IT R&D. These agencies report IT 

research budgets in the NITRD crosscut, 

and many other agencies with IT 

interests also participate informally in 

NITRD activities. The draft 2010 

Strategic Plan for the NITRD Program 

was developed by the NITRD agencies 

pursuant to a recommendation of the 

President’s Council of Advisors on 

Science and Technology (PCAST). 

Invitation to comment: Inputs of one 

page or less are welcomed in response 

to this third and final request for public 

comment on the Plan. E-mail to: nitrdsp@nitrd.gov. 

Submitted by the National Science 

Foundation for the National 

Coordination Office (NCO) for 

Networking and Information 

Technology Research and Development 

(NITRD) on September 1, 2010. 

September 13, 2010. 

Suzanne H. Plimpton, 

Reports Clearance Officer, National Science 

Foundation. 

[FR Doc. 2010–23459 Filed 9–20–10; 8:45 am] 

BILLING CODE 7555–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0297] 

Biweekly Notice Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

(the Act), the U.S. Nuclear Regulatory 

Commission (the Commission or NRC) 

is publishing this regular biweekly 

notice. The Act requires the 

Commission 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 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 August 26, 

2010, to September 8, 2010. The last 

biweekly notice was published on 

September 7, 2010 (75 FR 54390– 

54400). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.92, 

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 


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57522 Federal Register / Vol. 75, No. 182 / Tuesday, September 21, 2010 / Notices 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules, 

Announcements and Directives Branch 

(RADB), TWB–05–B01M, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be faxed to the RADB at 301–492– 

3446. Documents may be examined, 

and/or copied for a fee, at the NRC’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

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. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area O1F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/doc-collections/cfr/. If a 

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

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, any hearing held would 

take place before the issuance of any 

amendment. 

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 E–Filing rule 

(72 FR 49139, August 28, 2007). The E– 

Filing process requires participants to 

submit and serve all adjudicatory 

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 e-mail at 

hearing.docket@nrc.gov, or by telephone 

at (301) 415–1677, to request (1) a 

digital 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 

NRC’s public Web site at http:// 

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

apply-certificates.html. System 

requirements for accessing the E– 

Submittal server are detailed in NRC’s 

‘‘Guidance for Electronic Submission,’’ 

which is available on the agency’s 

public Web site at http://www.nrc.gov/ 

site-help/e-submittals.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 EIE, users will be 

required to install a Web browser plugVerDate Mar<15>2010 19:02 Sep 20, 2010 Jkt 220001 PO 00000 Frm 00086 Fmt 4703 Sfmt 4703 E:\FR\FM\21SEN1.SGM 21SEN1

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Federal Register / Vol. 75, No. 182 / Tuesday, September 21, 2010 / Notices 57523 

in from the NRC Web site. Further 

information on the Web-based 

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 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 e-mail notice 

confirming receipt of the document. The 

E–Filing system also distributes an email notice that provides access to the 

document to the NRC 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 agency’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 Web site at http:// 

www.nrc.gov/site-help/esubmittals.html, by e-mail at 

MSHD.Resource@nrc.gov, or by a tollfree call at (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 NRC’s 

electronic hearing docket which is 

available to the public at http:// 

ehd.nrc.gov/EHD_Proceeding/home.asp, 

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. 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. Nontimely filings will not be entertained 

absent a determination by the presiding 

officer that the petition or request 

should be granted or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

For further details with respect to this 

license amendment application, see the 

application for amendment which is 

available for public inspection at the 

Commission’s PDR, located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. Publicly 

available records will be accessible from 

the ADAMS Public Electronic Reading 

Room on the Internet at the NRC Web 

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

adams.html. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: July 20, 

2010. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

Limiting Condition for Operation (LCO) 

3.7.1.2, ‘‘Emergency Feedwater System,’’ 

to clarify the acceptability of 

transitioning from Mode 4 to Mode 3 

with the turbine-driven emergency 

feedwater (EFW) pump inoperable but 

available. This proposal would grant an 

exception to TS LCO 3.0.4 and 

Surveillance Requirement 4.0.4 

allowing entry into operational Mode 3 

with TS LCO equipment, the turbinedriven EFW pump, associated with a 

shutdown action inoperable. 

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 addition of an exception to 

TS LCO 3.0.4 for entry into Mode 3 during 

a plant startup for the turbine-driven EFW 

pump for a plant condition when the turbine 

driven EFW pump would be unable to 

complete its post maintenance activities (i.e. 

dynamic final calibration of the governor 

valve speed control unit governor control 

system) due to insufficient steam pressure in 

the steam generator secondary side and then 

to complete the quarterly IST [Inservice 

Testing] and 18 month EFAS [Engineered 

Safety Features Actuation System] SR 

[Surveillance Requirement] within the 

allowance of the delay of the respective SR 

is administrative in nature. 

This change will clarify that the turbinedriven EFW pump is not required to fully 

demonstrate operability (i.e. be inoperable 

pending completion of the quarterly IST and 

18 month EFAS SR) during plant startup 

prior to entry into Mode 3 under the 

conditions and for the period as provided in 

the quarterly IST and 18 month EFAS SR as 

granted by the NRC [Nuclear Regulatory 

Commission] in Reference 7.1 [NRC letter to 

Waterford 3 dated October 4, 2001, Waterford 

Steam Electric Station—Unit 3, Issuance of 

Amendment RE: Emergency Feedwater 

System (TAC No MB2010), Agencywide 

Documents Access and Management System 

(ADAMS) Accession No. ML012840538]. 

When the plant enters Mode 3 during plant 


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startup, the turbine-driven EFW pump is 

available (i.e., there is a reasonable 

expectation that once sufficient steam 

pressure is available to the turbine-driven 

EFW pump turbine, it will be able to 

successfully complete the quarterly IST and 

18 month EFAS surveillance requirements to 

fully demonstrate operability). 

Prior to entry into Mode 2, surveillance 

requirement testing of various combinations 

of EFW pumps and valves will ensure ALL 

required EFW system flow paths and 

equipment (which includes the turbinedriven EFW pump) are demonstrated 

operable before sufficient core heat is 

generated that would require the operation of 

the EFW System during a subsequent 

shutdown. 

Since the two motor-driven EFW pumps 

are required to be operable when entering 

Modes 3 from Mode 4, then for the worst case 

postulated accident scenario during plant 

startup, with the turbine-driven EFW pump 

considered inoperable but available (utilizing 

the exception to TS LCO 3.0.4 as tied to the 

quarterly IST and 18 month EFAS SR for 

fully demonstrating operability of the 

turbine-driven EFW pump), the EFW System 

safety function of achieving shutdown 

cooling entry conditions would be met. 

This request is merely a clarification and 

does not present any change to equipment 

operation, design or practices. The proposed 

clarification is not an accident initiator and 

will not adversely affect plant safety 

functions. The EFW System capability to 

provide its specified function of being able to 

achieve shutdown cooling entry conditions 

of the Reactor Coolant [S]ystem is unchanged 

by this clarification. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed addition of an exception to 

TS LCO 3.0.4 for entry into Mode 3 during 

a plant startup for the turbine-driven EFW 

pump for a plant condition when the turbinedriven EFW pump would be unable to 

complete its post maintenance activities (i.e. 

dynamic final calibration of the governor 

valve speed control unit governor control 

system) due to insufficient steam pressure in 

the steam generator secondary side and then 

to complete the quarterly IST and 18 month 

EFAS SR within the allowance of the delay 

of the respective SR is administrative in 

nature. 

This change will clarify that the turbinedriven EFW pump is not required to fully 

demonstrate operability (i.e. be inoperable 

pending completion of the quarterly IST and 

18 month EFAS SR) during plant startup 

prior to entry into Mode 3 under the 

conditions and for the period as provided in 

the quarterly IST and 18 month EFAS SR as 

granted by the NRC in Reference 7.1. When 

the plant enters Mode 3 during plant startup, 

the turbine-driven EFW pump is available 

(i.e. there is a reasonable expectation that 

once sufficient steam pressure is available to 

the turbine-driven EFW pump turbine, it will 

be able to successfully complete the quarterly 

IST and 18 month EFAS surveillance 

requirements to fully demonstrate 

operability). 

Prior to entry into Mode 2, surveillance 

requirement testing of various combinations 

of EFW pumps and valves will ensure ALL 

required EFW system flow paths and 

equipment (which includes the turbinedriven EFW pump) are demonstrated 

operable before sufficient core heat is 

generated that would require the operation of 

the EFW System during a subsequent 

shutdown. 

The addition of this exception to TS LCO 

3.0.4 for the turbine-driven EFW pump 

introduces no new mode of plant operation 

and does not alter the EFW System 

functional capability. The scope of this 

proposed change does not establish a 

potential new accident precursor. This 

proposed change will not change the design, 

configuration or method of operation of the 

EFW System. No new possibility for an 

accident is introduced by the proposed 

clarification. 

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 addition of an exception to 

TS LCO 3.0.4 for entry into Mode 3 during 

a plant startup for the turbine-driven EFW 

pump for a plant condition when the turbinedriven EFW pump would be unable to 

complete its post maintenance activities (i.e. 

dynamic final calibration of the governor 

valve speed control unit governor control 

system) due to insufficient steam pressure in 

the steam generator secondary side and then 

to complete the quarterly IST and 18 month 

EFAS SR within the allowance of the delay 

of the respective SR is administrative in 

nature. 

This change will clarify that the turbinedriven EFW pump is not required to fully 

demonstrate operability (i.e. be inoperable 

pending completion of the quarterly IST and 

18 month EFAS SR) during plant startup 

when entering Mode 3 under the conditions 

and for the period as provided in the 

quarterly IST and 18 month EFAS SR as 

granted by the NRC in Reference 7.1. When 

the plant enters Mode 3 during plant startup, 

the turbine-driven EFW pump is available 

(i.e. there is a reasonable expectation that 

once sufficient steam pressure is available to 

the turbine-driven EFW pump turbine, it will 

be able to successfully complete the quarterly 

IST and 18 month EFAS surveillance 

requirements to fully demonstrate 

operability). 

Prior to entry into Mode 2, surveillance 

requirement testing of various combinations 

of EFW pumps and valves will ensure ALL 

required EFW system flow paths and 

equipment (which includes the turbinedriven EFW pump) are demonstrated 

operable before sufficient core heat is 

generated that would require the operation of 

the EFW System during a subsequent 

shutdown. 

The proposed clarification does not 

adversely affect Emergency Feedwater 

equipment operating practices. The EFW 

System has the same capabilities as before to 

mitigate accidents. Surveillance requirements 

are not reduced by the proposed change. The 

EFW System capability to provide its 

specified function of being able to achieve 

shutdown cooling entry conditions of the 

Reactor Coolant System following a worst 

case postulated accident is unchanged by this 

clarification. 

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: Michael T. 

Markley. 

NextEra Energy Point Beach, LLC (the 

licensee), Docket Nos. 50–266 and 50– 

301, Point Beach Nuclear Plant (PBNP), 

Units 1 and 2, Town of Two Creeks, 

Manitowac County, Wisconsin 

Date of amendment request: April 7, 

2009, as supplemented by letters dated 

June 17, September 11, November 20, 

November 30, and December 8 of 2009; 

and February 11, February 25, April 22, 

April 30, July 21, July 28, and August 

2 of 2010. 

Description of amendment request: 

The proposed amendment would revise 

Reactor Protection System (RPS) and 

Engineered Safety Feature Actuation 

System (ESFAS) instrumentation 

setpoints for the PBNP, Units 1 and 2. 

The revised Technical Specification 

(TS) allowable values are specified in 

Tables 3.3.1–1 and 3.3.2–1 for RPS and 

ESFAS, respectively. These changes 

were originally included as part of the 

April 7, 2009, extended power uprate 

(EPU) license amendment request, but 

subsequently divided into a separate 

licensing action for independent 

technical review. The proposed changes 

include both EPU and non-EPU related 

changes. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 


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consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to the TSs will 

ensure that the results of previously 

evaluated accidents at the uprated conditions 

remain within the acceptance criteria. The 

proposed RPS and ESFAS setpoint changes 

provide appropriate values for operation at 

EPU conditions. The revised TS allowable 

values have been calculated to account for 

new EPU analytical limits, instrument 

uncertainties, and instrument drift. The 

proposed RPS and ESFAS setpoint changes 

are considered in the safety analysis for the 

affected RPS and ESFAS functions, and do 

not significantly increase the probability or 

consequences of the accidents previously 

evaluated and the setpoint changes 

considered in the safety analysis continue to 

meet the applicable acceptance criteria. The 

safety analyses for these accidents have been 

performed at the EPU power level and 

demonstrated acceptable results. 

The proposed changes will ensure that the 

instruments actuate as assumed to mitigate 

accidents previously evaluated. The 

proposed changes will not significantly affect 

accident initiators or precursors and will not 

alter or prevent the ability of systems, 

structures, or components from performing 

the intended safety function to meet the 

applicable acceptance limits for the accidents 

and events. 

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 change does not involve a physical 

alteration of the plant or change the methods 

governing normal plant operation. The 

change does not alter assumptions made in 

the safety analyses, but ensures that the 

instruments behave as assumed in the 

accident analysis. The proposed change is 

consistent with the safety analysis 

assumptions. The proposed RPS and ESFAS 

Limiting Safety System Setting (LSSS) 

changes do not create the possibility of a new 

or different type of accident due to operation 

at EPU conditions. The revised TS LSSS 

values have been calculated to account for 

new EPU analytical limits and known 

instrument uncertainties. The proposed RPS 

and ESFAS setpoint changes are used in the 

safety analysis for the affected RPS and 

ESFAS functions, and do not significantly 

affect these accidents or the applicable 

acceptance criteria. 

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 changes clarify the TS 

requirements for instrumentation to ensure 

that the automatic protection action will 

correct the abnormal situation before a safety 

limit is exceeded. The proposed change also 

revises the TSs to enhance the controls used 

to maintain the variables and systems within 

the prescribed operating ranges, in order to 

ensure that automatic protection actions 

occur to initiate the operation of systems and 

components important to safety as assumed 

in the accident analysis. No change is made 

to the accident analysis assumptions. 

The proposed changes to the RPS and 

ESFAS setpoint TSs provide adequate margin 

such that PBNP Units 1 and 2 can be 

operated in a safe manner at EPU conditions. 

No new accident scenarios, failure 

mechanisms, or single failures are introduced 

as a result of the proposed changes. All 

systems, structures and components 

previously assumed for the mitigation of an 

event remain capable of fulfilling their 

intended function. The proposed changes 

will not have any significant effect on the 

margin of safety. 

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: William Blair, 

Senior Attorney, NextEra Energy Point 

Beach, LLC, P. O. Box 14000, Juno 

Beach, FL 33408–0420. 

NRC Branch Chief: Robert J. 

Pascarelli. 

NextEra Energy Point Beach, LLC (the 

licensee), Docket Nos. 50–266 and 50– 

301, Point Beach Nuclear Plant, Units 1 

and 2, Town of Two Creeks, Manitowac 

County, Wisconsin 

Date of amendment request: April 7, 

2009, as supplemented by letters dated 

June 17 (two letters), September 11, 

September 25, October 9, November 20 

(two letters), November 21 (two letters), 

November 30, December 8, and 

December 16 of 2009; and January 7, 

January 8, January 22, February 11, 

February 25, March 3, April 15, April 

22, July 8, July 28, August 2, August 9, 

and August 24 of 2010. 

Description of amendment request: 

The proposed amendment would 

change the auxiliary feedwater (AFW) 

system design and Technical 

Specifications (TS) 3.7.5, ‘‘Auxiliary 

Feedwater (AFW),’’ and TS 3.7.6, 

‘‘Condensate Storage Tank (CST),’’ 

resulting from (1) modifications to the 

AFW system to support requirements 

for transients and other accidents at 

extended power uprate (EPU) 

conditions; (2) installation of main 

feedwater isolation valves to support 

accident mitigation by ensuring that 

containment pressure does not exceed 

safety analysis limits; (3) automatic 

AFW switchover from a CST suction 

source to a safety-related Service Water 

(SW) source; and (4) setpoint changes 

supporting the aforementioned physical 

modifications. These changes were 

originally included as part of the April 

7, 2009, EPU license amendment 

request, but subsequently divided into a 

separate licensing action for 

independent technical review. The 

upgrades and modifications to the AFW 

system are being installed to provide 

additional capacity and reliability for 

the system. Although the proposed 

changes are also designed to support the 

requirements for transients and other 

accidents at EPU conditions, the 

proposed changes for this amendment 

are being evaluated using the current 

licensing basis. 

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 NRC staff performed 

its own analysis, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The design functions of the AFW system 

will not be altered by the proposed change. 

The AFW system will continue to perform its 

original intended design function, mitigating 

the consequences of accidents previously 

evaluated. The proposed changes will not 

significantly affect accident initiators or 

precursors. No new accident scenarios, 

failure mechanisms, or single failures are 

introduced as a result of the proposed 

modifications. 

Implementation of the new AFW system 

design and the proposed changes to TS 3.7.5 

was evaluated against the current analysis of 

record for the current licensed power level at 

PBNP, Units 1 and 2. The current analyses 

remain applicable or are unaffected by 

implementation of the new AFW system and 

associated TS changes, with the exception of 

the steam line break containment response 

and steam generator tube rupture (SGTR) 

radiological consequences. These two 

accidents were reanalyzed with the current 

licensing basis for the AFW modifications 

and the results were acceptable with the 

revised minimum and maximum AFW flow 

rates and pump start timing. 

Therefore, the consequences of accidents 

previously evaluated for the current licensed 

power level are not significantly increased. 

A proposed change to TS 3.7.6 changes the 

surveillance requirement (SR) for minimum 

CST water inventory to be maintained to 

supply AFW pump suction in the event of a 

Station Blackout, when the safety-related 

AFW suction source from the SW system is 

not available. The proposed TS 3.7.6 SR 

increases the current minimum required 

inventory to account for the increased flow 

rates from the new AFW system design, 


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suction piping losses, instrument 

uncertainties, vortex prevention, net positive 

suction head (NPSH) requirements, and the 

suction of the AFW pumps under various 

combinations of CST and plant units in 

operation. This change to the minimum 

required CST level inventory will not 

increase the probability or consequences of 

previously evaluated accidents. 

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 introduce a 

new mode of plant operation. The proposed 

changes involving the AFW system do not 

significantly alter any design basis accident 

or event response. The proposed changes will 

not significantly affect accident initiators or 

precursors. The AFW system will continue to 

perform its design function. No new accident 

scenarios, failure mechanisms, or single 

failures are introduced as a result of the 

proposed modifications. All systems, 

structures, and components previously 

assumed for the mitigation of an event 

remain capable of fulfilling their intended 

design function. The new AFW system 

design and proposed changes to TS 3.7.5 and 

the proposed increase in CST inventory in TS 

3.7.6 do not create the possibility of a new 

or different kind of accident or event. 

As previously discussed, implementation 

of the new AFW system design and the 

proposed changes to TS 3.7.5 was evaluated 

against the current analysis of record for the 

current licensed power level at PBNP, Units 

1 and 2. The current analyses remain 

applicable or are unaffected by 

implementation of the new AFW system and 

associated TS changes, with the exception of 

the steam line break containment response 

and steam generator tube rupture (SGTR) 

radiological consequences. These two 

accidents were reanalyzed with the current 

licensing basis for the AFW modifications 

and the results are acceptable with the 

revised minimum and maximum AFW flow 

rates and pump start timing. The AFW 

system design change, the changes to TS 

.3.7.5, and the increase in required CST 

inventory established in TS 3.7.6, are not 

significant accident initiators or precursor 

and will not create the possibility of a new 

or different kind of accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The upgrade to the AFW system is being 

made to support requirements for transients 

and other accidents at EPU conditions. This 

modification to the AFW system will provide 

additional capacity and reliability for the 

system. As such, the proposed amendment 

does not involve a significant reduction in 

safety. 

The analyses and evaluations of the 

Nuclear Steam Supply System (NSSS) and 

Balance of Plant (BOP) systems based on 

completion of the required modifications, 

confirm that the systems and components 

will function as designed and demonstrate 

that the NSSS and BOP systems and 

components meet all applicable design and 

licensing requirements at the uprated power 

level. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

Based on the above review, it appears 

that the three standards of 10 CFR 

50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William Blair, 

Senior Attorney, NextEra Energy Point 

Beach, LLC,.P. O. Box 14000, Juno 

Beach, FL 33408–0420. 

NRC Branch Chief: Robert J. 

Pascarelli. 

NextEra Energy Point Beach, LLC (the 

licensee), Docket Nos. 50–266 and 50– 

301, Point Beach Nuclear Plant, Units 1 

and 2, Town of Two Creeks, Manitowac 

County, Wisconsin 

Date of amendment request: June 1, 

2010, as supplemented by letter dated 

July 9, 2010. 

Description of amendment request: 

The proposed amendment consists of 

revising the current license basis 

regarding a postulated reactor vessel 

head (RVH) drop event to conform to 

the NRC-endorsed guidance of Nuclear 

Energy Institute (NEI) 08–05, ‘‘Industry 

Initiative on Control of Heavy Loads,’’ 

Revision 0. The proposed change to the 

license basis will revise Chapter 14.3.6, 

‘‘Reactor Vessel Head Drop Event,’’ of 

the Final Safety Analysis Report. The 

current license basis assumes failure of 

the reactor coolant system (RCS) 

boundary caused by the predicted 

maximum downward displacement of 

the reactor vessel which would sever all 

36 bottom-mounted instrument (BMI) 

conduit tubes. The new analysis 

demonstrates that a postulated RVH 

drop would not result in a loss of RCS 

inventory caused by an RCS boundary 

failure, since the BMI conduits would 

remain intact. 

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 is limited in 

scope to a postulated RVH drop and the 

administrative controls in place, which limit 

the height of the RVH lift, ensuring an actual 

drop is bounded by the analyses of record. 

Incorporation of the analysis performed in 

accordance with NRC-approved guidance, 

which demonstrates bottom-mounted 

instrumentation (BMI) conduits will not 

sever following a postulated RVH drop, does 

not increase the probability or consequences 

of a previously evaluated accident. The 

evaluation, in fact, demonstrates that if the 

postulated RVH drop occurred, the 

consequences would be significantly less 

than are now assumed because the ability to 

maintain a coolable geometry in the core has 

not been compromised. In accordance with 

NRC-endorsed methodology contained in NEI 

08–05, which states, ‘‘Previous evaluations 

have indicated that the consequences of 

impacts between the upper vessel internals 

and the fuel were not significant with respect 

to public health and safety,’’ a revised 

radiological analysis was not performed. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment is limited in 

scope to a postulated RVH drop and the 

administrative controls in place, which limit 

the height of the reactor RVH lift, ensuring 

an actual drop is bounded by the analysis of 

record. 

Incorporation of the analysis performed in 

accordance with NRC-approved guidance, 

which demonstrates BMI conduits will not 

sever following a postulated RVH drop, does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. The proposed 

amendment does not: (1) Operate equipment 

in alignments or in a manner different form 

that previously evaluated in the FSAR; (2) 

install, remove or modify equipment 

important to safety; or (3) introduce new 

failure modes or effects for any existing 

system, structure or component. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of any accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment is limited in 

scope to a postulated RVH drop and the 

administrative controls in place, which limit 

the height of the reactor RVH lift, ensuring 

an actual drop is bounded by the analysis of 

record. 

Incorporation of the analysis performed in 

accordance with NRC-approved guidance, 

which demonstrates BMI conduits will not 

sever following a postulated RVH drop, does 

not involve a significant reduction in the 

margin of safety. The evaluation, in fact, 

demonstrates that if the postulated RVH drop 

occurred, the consequences would be 

significantly less than are now assumed 

because the ability to maintain a coolable 

geometry in the core has not been 

compromised. 


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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: William Blair, 

Senior Attorney, NextEra Energy Point 

Beach, LLC, P. O. Box 14000, Juno 

Beach, FL 33408–0420. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Northern States Power Company— 

Minnesota, Docket No. 50–263, 

Monticello Nuclear Generating Plant 

(MNGP), Wright County, Minnesota 

Date of amendment request: January 

21, 2010. 

Description of amendment request: 

The licensee proposed to amend the 

MNGP Technical Specifications to allow 

operation in the Maximum Extended 

Load Line Limit Analysis Plus 

(MELLLA+) expanded domain. The 

licensee stated that the Nuclear 

Regulatory Commission (NRC) had 

previously approved various aspects of 

the MELLLA+ methodology, but that the 

current application is the first plantspecific use of such methodology. The 

amendment would include changes to 

the Technical Specifications to: (1) 

Prohibit the use of the MELLLA+ 

expanded operating domain when in 

single loop operation; (2) change the 

allowable value for Average Power 

Range Monitor (APRM)-Simulated 

Thermal Power—High; (3) eliminate an 

unnecessary surveillance requirement; 

(4) require certain content in the Core 

Operating Limits Report. Approval of 

this amendment would allow the 

licensee to implement operational 

changes to provide increased 

operational flexibility for power 

maneuvering, to compensate for fuel 

depletion, and to maintain efficient 

power distribution in the reactor core 

without the need for more frequent rod 

pattern changes. MELLLA+ would 

increase the operating range to the 

Extended Power Uprate rated thermal 

power at 80 percent flow; thus creating 

a 20 percent flow-control window. By 

operating in the MELLLA+ domain, a 

significantly lower number of control 

rod movements will be required than in 

the present operating domain. This 

would represent a significant 

improvement in operating flexibility. It 

also provides safer operation, because 

reducing the number of control rod 

manipulations would minimize the 

likelihood of fuel failures, and reduce 

the likelihood of accidents initiated by 

reactor maneuvers. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (10 CFR) Part 

50.91(a), the licensee has provided its 

analysis of the issue of no significant 

hazards consideration (NSHC). The 

licensee’s NSHC analysis is reproduced 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The probability (frequency of occurrence) 

of [d]esign [b]asis [a]ccidents occurring is not 

affected by the MELLLA+ operating domain, 

because MNGP continues to comply with the 

regulatory and design basis criteria 

established for plant equipment. Further, a 

probabilistic risk assessment demonstrates 

that the calculated core damage frequencies 

do not significantly change due to the 

MELLLA+. 

There is no change in consequences of 

postulated accidents, when operating in the 

MELLLA+ operating domain compared to the 

operating domain previously evaluated. The 

results of accident evaluations remain within 

the NRC[-]approved acceptance limits. 

The spectrum of postulated transients has 

been investigated and is shown to meet the 

plant’s currently licensed regulatory criteria. 

In the area of fuel and core design, for 

example, the Safety Limit Minimum Critical 

Power Ratio (SLMCPR) is still met. 

Continued compliance with the SLMCPR 

will be confirmed on a cycle[-]specific basis 

consistent with the criteria accepted by the 

NRC. 

Challenges to the [r]eactor [c]oolant 

[p]ressure [b]oundary were evaluated for the 

MELLLA+ operating domain conditions 

(pressure, temperature, flow, and radiation) 

and were found to meet their acceptance 

criteria for allowable stresses and 

overpressure margin. 

Challenges to the containment were 

evaluated and the containment and its 

associated cooling systems continue to meet 

the current licensing basis. The calculated 

post[-]LOCA [loss-of-coolant accident] 

suppression pool temperature remains 

acceptable. 

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. 

Equipment that could be affected by the 

MELLLA+ operating domain has been 

evaluated. No new operating mode, safetyrelated equipment lineup, accident scenario, 

or equipment failure mode was identified. 

The full spectrum of accident considerations 

has been evaluated and no new or different 

kind of accident has been identified. The 

MELLLA+ operating domain uses developed 

technology and applies it within the 

capabilities of existing plant safety-related 

equipment in accordance with the regulatory 

criteria (including NRC approved codes, 

standards and methods). No new accident or 

event precursor has been identified. 

The-MNGP TS require revision to 

implement the MELLLA+ operating domain. 

The revisions have been assessed and it was 

determined that the proposed change will not 

introduce a different accident than that 

previously evaluated. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The MELLLA+ operating domain affects 

only design and operational margins. 

Challenges to the fuel, reactor coolant 

pressure boundary, and containment were 

evaluated for the MELLLA+ operating 

domain conditions. Fuel integrity is 

maintained by meeting existing design and 

regulatory limits. The calculated loads on 

affected structures, systems and components, 

including the reactor coolant pressure 

boundary, will remain within their design 

allowables for design[-]basis event categories. 

No NRC acceptance criterion is exceeded. 

Because the MNGP configuration and 

responses to transients and postulated 

accidents do not result in exceeding the 

presently approved NRC acceptance’ limits, 

the proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the proposed 

amendment involves no significant 

hazards consideration. 

Attorney for the licensee: Peter M. 

Glass, Assistant General Counsel, Xcel 

Energy Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant, Units 1 and 2, Goodhue County, 

Minnesota 

Date of amendment request: June 14, 

2010. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specifications to 

allow the use of a dedicated on-line core 

power distribution monitoring system 

(PDMS) to enhance surveillance of core 

thermal limits. The PDMS to be used at 

Prairie Island Nuclear Generating Plant, 

Units 1 and 2, is the Westinghouse 

proprietary core analysis system called 

the Best Estimate Analyzer for Core 

Operations—Nuclear (BEACONTM). 


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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 PDMS performs continuous core 

power distribution monitoring with data 

input from existing plant instrumentation. 

The system passively supports Technical 

Specification (TS) surveillances which 

ensure that core power distribution is within 

the same limits that are currently prescribed. 

Further, the proposed TS Actions are 

comparable to existing operator actions such 

that no new plant configurations are 

prompted by the proposed change. The 

system’s physical interface with plant 

equipment is limited to an electronic link 

from a new workstation to the plant process 

computer. The system is passive in that it 

provides no control or alarm functions, and 

does not promote any new plant 

configuration which would affect the 

initiation, probability, or consequences of a 

previously-evaluated accident. Continuous 

on-line core monitoring through the use of 

PDMS provides significantly more 

information about the power distributions 

present in the core than is currently 

available. This system performance may 

result in an earlier determination of an 

adverse core condition and more time for 

operator action, thus reducing the probability 

of an accident occurrence and reduced 

consequences should a previously-evaluated 

accident occur. 

By virtue of its inherently passive 

surveillance function and limited interface 

with plant systems, structures, or 

components, the proposed changes will not 

result in any additional challenges to plant 

equipment that could increase the probability 

or occurrence of any previously-evaluated 

accident. Further, the proposed changes will 

ensure conformance to the same core power 

distribution limits that form the basis for 

initial conditions of previously evaluated 

accidents. Thereby, the proposed changes 

will not affect the consequences of any 

previously-evaluated accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequence of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The system’s physical interface with plant 

equipment is limited to an electronic link 

from a new workstation to the plant process 

computer. The system is passive in that it 

provides no control or alarm functions, and 

the proposed changes (including operator 

actions prescribed by the proposed TS) do 

not promote any new plant configuration 

which would create the possibility for an 

accident of a new or different type. 

The NRC previously evaluated the effects 

of using the PDMS to monitor core power 

distribution parameters and determined that 

all design standards and applicable safety 

criteria limits are met. The Technical 

Specifications will continue to require 

operation within the required core operating 

limits, and appropriate actions will continue 

to be taken when or if limits are exceeded. 

Thus, the reactor core will continue to be 

operated within its reference bounds of 

design such that an accident of a new or 

different type is not credible. 

The proposed change, therefore, does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

No margin of safety is adversely affected by 

the implementation of the PDMS. The 

margins of safety provided by current TS 

requirements and limits remain unchanged, 

as the TS will continue to require operation 

within the core limits that are based on NRCapproved reload design methodologies. The 

proposed change does not result in changes 

to the core operating limits. Appropriate 

measures exist to control the values of these 

cycle-specific limits, and appropriate actions 

will continue to be specified and taken when 

limits are violated. Such actions remain 

unchanged. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Robert J. 

Pascarelli. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: May 18, 

2010. 

Description of amendment request: 

The proposed amendments would 

reduce system/equipment diversity in 

isolation of low-pressure residual heat 

removal (RHR) system from highpressure reactor coolant system (RCS). 

The change will allow similarly 

qualified pressure transmitters to be 

used in more than one RHR train as 

necessary regardless of manufacturer of 

the transmitters. 

The valves separating the RHR from 

the RCS are to have independent and 

diverse interlocks to prevent both from 

opening unless the RCS pressure is 

below that of the RHR in compliance 

with the Nuclear Regulatory 

Commission’s Technical Position ICSB– 

3, ‘‘Isolation of Low Pressure Systems 

from the High Pressure Reactor Coolant 

System.’’ Consequently, the change 

would result in more than minimal 

increase in the likelihood of a 

malfunction of systems, structures, or 

components important to safety as 

previously evaluated in the plants’ 

Updated Final Safety Analysis Report. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revising the 

justification for diversity associated with the 

RHR isolation valves will not cause an 

accident to occur and will not result in any 

change in the operation of the associated 

accident mitigation equipment. The proposed 

changes will not revise the operability 

requirements (e.g., leakage limits) for the 

RHR system. The design-basis accidents will 

remain the same postulated events described 

in the STP Unit 1 and Unit 2 Updated Final 

Safety Analysis Report[,] and the 

consequences of the design-basis accidents 

will remain the same. 

Therefore, the proposed changes will not 

increase the probability or consequences of 

an accident previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes will not alter the 

plant configuration or require any unusual 

operator actions. The proposed changes will 

not alter the way any structure, system, or 

component functions, and will not 

significantly alter the manner in which the 

plant is operated. The response of the plant 

and the operators following an accident will 

not be different. In addition, the proposed 

changes do not introduce any new failure 

modes. In the event the RHR system is 

overpressurized by the RCS, all leakages 

originating from RHR components will be 

detected by the Reactor Coolant Pressure 

Boundary Leakage Detection System as 

discussed in the STP UFSAR [Updated Final 

Safety Analysis Report]. 

Therefore, the proposed changes will not 

create the possibility of a new or different 

kind of accident from any accident 

previously analyzed. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change to revise the 

rationale for diversity associated with RHR 

system isolation valve operation will not 


srobinson on DSKHWCL6B1PROD with NOTICES
Federal Register / Vol. 75, No. 182 / Tuesday, September 21, 2010 / Notices 57529 

cause an accident to occur and will not result 

in any change in the operation of the 

associated accident mitigation equipment. 

The operability requirements for the isolation 

valves have not been changed, and the RHR 

system will continue to function as assumed 

in the safety analysis. In addition, the 

proposed changes will not adversely affect 

equipment design or operation, and there are 

no changes being made to required safety 

limits or safety system settings that would 

adversely affect plant safety. 

Therefore, the proposed changes will not 

result in a reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendments involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: May 18, 

2010. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specification (TS) 

6.8.3.l, ‘‘Containment Post-Tensioning 

System Surveillance Program.’’ TS 

6.8.3.l states that the containment posttensioning system surveillance program 

shall be in accordance with American 

Society of Mechanical Engineers 

(ASME) Code, Section XI, Subsection 

IML, 1992 Edition with 1992 Addenda, 

as supplemented by 10 CFR 

50.55a(b)(2)(viii). The current 

inspection interval of South Texas 

Project (STP), Units 1 and 2 ends in 

September 2010. The proposed 

amendments will provide for updating 

the surveillance program consistent 

with the updated edition of the ASME 

Code, Section XI as required by 10 CFR 

50.55a. 

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 Technical Specification 

change removes the specific edition of the 

ASME [C]ode to be applied. Inspection 

practices will continue to be consistent with 

the approved ASME [C]ode edition. The 

proposed change is consistent with NUREG– 

1481 [guidance]. 

Therefore, the proposed changes will not 

increase the probability or consequences of 

an accident previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes will not alter the 

plant configuration (no new or different type 

of equipment will be installed) or require any 

unusual operator actions. The proposed 

changes will not alter the way any structure, 

system, or component functions, and will not 

significantly alter the manner in which the 

plant is operated. The response of the plant 

and the operators following an accident will 

not be different. In addition, the proposed 

change does not introduce any new failure 

modes. 

Therefore, the proposed changes will not 

create the possibility of a new or different 

kind of accident from any accident 

previously analyzed. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed Technical Specification 

change removes the specific edition of the 

ASME [C]ode to be applied. Inspection 

practices will continue to be consistent with 

the approved ASME [C]ode edition. The 

change is consistent with NUREG–1481 

guidance. 

Therefore, the proposed changes will not 

result in a reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendments involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: June 28, 

2010. 

Description of amendment requests: 

The proposed amendments request 

correction of an oversight in previous 

amendments (Amendment No. 185 to 

Facility Operating License No. NPF–76 

and Amendment No. 172 to Facility 

Operating License No. NPF–80) that 

revised the Technical Specifications 

(TSs) regarding control room envelope 

(CRE) habitability in accordance with 

TS Task Force (TSTF) Traveler No. 448, 

Revision 3. In its application for those 

previous amendments, STP Nuclear 

Operating Company (STPNOC) did not 

specify what shutdown actions would 

be taken if required actions for an 

inoperable CRE boundary were not met. 

This was inconsistent with TSTF–448. 

The proposed amendments would 

correct this oversight. STPNOC also 

requested to add a note to the required 

actions for inoperable CRE boundary to 

clarify that the boundary is not a 

required system, subsystem, train, 

component, or device that depends on 

a diesel generator as a source of 

emergency power. This change would 

clarify the application of TS action 

3.8.1.1, ‘‘AC Sources, DC Sources, and 

Other Power Distribution,’’ when the 

CRE is inoperable. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to add the shutdown 

actions to TS ACTION 3.7.7.d is consistent 

with Nuclear Regulatory Commission (NRC) 

noticed Industry/Technical Specification 

Task Force (TSTF) Standard Technical 

Specification (STS) change TSTF–448 

Revision 3, which has been approved by an 

NRC safety evaluation. 

The proposed change to add a note to the 

required action for an inoperable control 

room envelope boundary does not change the 

design function of the Control Room Makeup 

and Cleanup Filtration Systems or the design 

function of the A.C. Sources, D.C. Sources, 

and Onsite Power Systems or how these 

systems operate. The change only clarifies 

that the Control Room Envelope boundary is 

not a required system, subsystem, train, 

component, or device that depends on a 

diesel generator as a source of emergency 

power. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to add the shutdown 

actions to TS ACTION 3.7.7.d is consistent 

with Nuclear Regulatory Commission (NRC) 

noticed Industry/Technical Specification 

Task Force (TSTF) Standard Technical 

Specification (STS) change TSTF–448 

Revision 3, which has been approved by an 

NRC safety evaluation. 

The proposed change to add a note to the 

required action for an inoperable control 

room envelope boundary does not change the 

design of the Control Room Makeup and 


srobinson on DSKHWCL6B1PROD with NOTICES
57530 Federal Register / Vol. 75, No. 182 / Tuesday, September 21, 2010 / Notices 

Cleanup Filtration Systems or the design 

function of the A.C. Sources, D.C. Sources, 

and Onsite Power Systems. The change only 

clarifies that the Control Room Envelope 

boundary is not a required system, 

subsystem, train, component, or device that 

depends on a diesel generator as a source of 

emergency power. 

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 to a margin of safety? 

Response: No. 

The proposed change to add the shutdown 

actions to TS ACTION 3.7.7.d is consistent 

with Nuclear Regulatory Commission (NRC) 

noticed Industry/Technical Specification 

Task Force (TSTF) Standard Technical 

Specification (STS) change TSTF–448 

Revision 3, which has been approved by an 

NRC safety evaluation. 

The proposed change to add a note to the 

required action for an inoperable control 

room envelope boundary does not change 

any safety margins associated with operation 

of the Control Room Makeup and Cleanup 

Filtration Systems or any safety margins 

associated with the A.C. Sources, D.C. 

Sources, and Onsite Power Systems. The 

change only clarifies that the Control Room 

Envelope boundary is not a required system, 

subsystem, train, component, or device that 

depends on a diesel generator as a source of 

emergency power. 

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 standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendments involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

Notice of Issuance of Amendments to 

Facility Operating 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. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, 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 are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/adams.html. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of application of amendments: 

August 31, 2009, as supplemented April 

14, 2010. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications to allow one of the two 

required 230 kV switchyard 125 Vdc 

power sources (batteries) to be 

inoperable for up to 10 days for the 

purpose of replacing an entire battery 

bank and performing the required 

testing. 

Date of Issuance: August 30, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 370, 372, 371. 

Renewed Facility Operating License 

Nos. DPR–38, DPR–47, and DPR–55: 

Amendments revised the licenses and 

the technical specifications. 

Date of initial notice in Federal 

Register: March 9, 2010 (75 FR 10828). 

The supplement dated April 14, 2010, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 30, 

2010. 

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 (RBS), 

West Feliciana Parish, Louisiana 

Date of amendment request: August 

10, 2009, as supplemented by letters 

dated December 8, 2009, and April 22, 

June 16, and August 17, 2010, and by 

emails dated June 29, July 12, and July 

28, 2010. 

Brief description of amendment: The 

amendment revised the TSs for the RBS 

to support operation with 24-month fuel 

cycles. By letter dated June 16, 2010, 

Entergy withdrew its proposed changes 

to TS 3.3.8 regarding the change to the 

degraded voltage instrumentation 

allowable values as indicated on Table 

3.3.8.1–1 and to extend the Surveillance 

Requirement (SR) 3.3.8.1.3 and SR 

3.3.8.1.4 from 18 to 24 months. By letter 

dated August 17, 2010, Entergy 

withdrew the request for not revising SR 

3.3.8.1.4 and requested that this SR be 

extended as originally requested. 

Date of issuance: August 31, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 180 

days from the date of issuance. 

Amendment No.: 168. 

Facility Operating License No. NPF– 

47: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 20, 2009 (74 FR 

53776). 

The supplements dated December 8, 

2009, April 22, June 16, and August 17, 

2010, and emails dated June 29, July 12, 

and July 28, 2010, 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 August 31, 2010. 


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No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Nuclear 

Plant, Van Buren County, Michigan 

Date of application for amendment: 

August 25, 2009 supplemented by letter 

dated May 3, 2010. 

Brief description of amendment: The 

amendment modifies technical 

specification 5.5.14, ‘‘Containment 

Leakage Rate Testing Program,’’ to allow 

a one-time extension to the 10-year 

frequency for the next 10 CFR Part 50 

Appendix J, Option B, Type A, 

containment integrity leakage test 

(ILRT) or Type A test at Palisades 

Nuclear Plant. This amendment permits 

the existing ILRT frequency to be 

extended from 10 years (120 months) to 

approximately 11.25 years (135 

months). This amendment also prevents 

the necessity of performing a Type A 

test six months prior to the 10th 

anniversary of the completion of the last 

Type A test, which was completed on 

May 3, 2001. 

Date of issuance: August 23, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 240. 

Facility Operating License No. DPR– 

20: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 20, 2009 (74 FR 

53777). 

The supplemental letters contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. The 

Commission’s related evaluation of the 

amendment is contained in a Safety 

Evaluation dated August 23, 2010. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of application for amendments: 

October 23, 2008, as supplemented by 

letters dated September 28, and 

November 18, 2009, March 29, and 

August 3, 2010. 

Brief description of amendments: The 

amendments revise the Technical 

Specifications to support the 

application of alternative source term 

methodology with respect to the loss-ofcoolant accident and the fuel-handling 

accident. 

Date of issuance: September 6, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment Nos.: 197, 184. 

Facility Operating License Nos. NPF– 

11 and NPF–18: The amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: April 7, 2009 (74 FR 15771). 

The September 28, and November 18, 

2009, March 29, and August 3, 2010 

supplements contained clarifying 

information and did not change the NRC 

staff’s initial proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 6, 

2010. 

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 (PBAPS), Units 2 and 3, 

York and Lancaster Counties, 

Pennsylvania 

Date of application for amendments: 

August 31, 2009. 

Brief description of amendments: The 

amendments modify the PBAPS 

Technical Specifications (TS) by 

relocating specific surveillance 

frequencies to a licensee-controlled 

program with the implementation of 

Nuclear Energy Institute (NEI) 04–10, 

‘‘Risk-Informed Technical Specifications 

Initiative 5b, Risk-Informed Method for 

Control of Surveillance Frequencies.’’ 

Additionally, the change adds a new 

program, the Surveillance Frequency 

Control Program, to TS Section 5, 

Administrative Controls. The changes 

are based on NRC-approved Industry 

Technical Specifications Task Force 

(TSTF) Traveler 425, Revision 3, 

‘‘Relocate Surveillance Frequencies to 

Licensee Control—Risk Informed 

Technical Specification Task Force 

Initiative 5b,’’ with optional changes and 

variations as described in Attachment 1, 

Section 2.2 of the licensee’s submittal 

dated August 31, 2009. 

Date of issuance: August 27, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 278 and 281. 

Renewed Facility Operating License 

Nos. DPR–44 and DPR–56: Amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: May 5, 2010 (75 FR 23815). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 27, 

2010. 

No significant hazards consideration 

comments received: No. 

NextEra Energy Seabrook, LLC, Docket 

No. 50–443, Seabrook Station, Unit No. 

1, Rockingham County, New Hampshire 

Date of amendment request: March 

16, 2010, as supplemented on July 9, 

2010. 

Description of amendment request: 

This amendment revises the Seabrook 

Technical Specifications requirement 

that the Operations Manager shall have 

held a senior reactor operator license for 

the Seabrook Station prior to assuming 

the Operations Manager position. 

Specifically, the proposed change now 

requires the Operations Manager to meet 

one of the following: (1) Hold a senior 

operator license; (2) have held a senior 

operator license for a similar unit; or (3) 

have been certified for equivalent senior 

operator knowledge. 

Date of issuance: September 2, 2010. 

Effective date: As of its date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 124. 

Facility Operating License No. NPF– 

86: The amendment revised the TS and 

the License. 

Date of initial notice in Federal 

Register: May 4, 2010 (75 FR 23816). 

The supplemental letter dated July 9, 

2010, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 2, 

2010. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–311, 

Salem Nuclear Generating Station, Unit 

No. 2, Salem County, New Jersey 

Date of application for amendment: 

March 29, 2010, as supplemented on 

June 25, and August 18, 2010. 

Brief description of amendments: The 

amendment revises the Technical 

Specifications (TSs) to allow a one-time 

replacement of the 2C 125-volt direct 

current battery while Salem Unit No. 2 

is at power. 

Date of issuance: September 1, 2010. 

Effective date: As of the date of 

issuance, to be implemented within 30 

days. 

Amendment No.: 280. 


srobinson on DSKHWCL6B1PROD with NOTICES
57532 Federal Register / Vol. 75, No. 182 / Tuesday, September 21, 2010 / Notices 

Facility Operating License No. DPR– 

75: The amendment revised the TSs and 

the License. 

Date of initial notice in Federal 

Register: June 1, 2010 (75 FR 30446). 

The letters dated June 25, and August 

18, 2010, provided clarifying 

information that did not change the 

initial proposed no significant hazards 

consideration determination or expand 

the application beyond the scope of the 

original Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 1, 

2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 10th day 

of September 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–23388 Filed 9–20–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0301; EA–10–054] 

In the Matter of: Stone & Webster 

Construction, Inc.; Confirmatory Order 

(Effective Immediately) 

I 

Stone & Webster Construction, Inc. 

(SWCI), a Shaw Group company 

(referred to as Shaw), provides 

integrated services to various industries 

including the nuclear power industry. 

Shaw provides services to over thirty 

(30) operating nuclear plants and other 

facilities regulated by the U.S. Nuclear 

Regulatory Commission (NRC or 

Commission). 

This Confirmatory Order is the result 

of an agreement reached during an 

alternative dispute resolution (ADR) 

mediation session conducted on August 

24, 2010 in Washington, DC. 

II 

By letter dated June 2, 2010, the NRC 

identified to Shaw an apparent violation 

of 10 CFR. 50.7, ‘‘Employee Protection,’’ 

relating to the termination of a former 

painter foreman in May 2004 at the 

Browns Ferry Nuclear Power Plant. The 

apparent violation was issued based on 

the U.S. Department Labor (DOL) 

Administrative Review Board’s (ARB) 

September 24, 2009 decision (ARB Case 

No. 06–041). The ARB reversed a 

January 9, 2006, DOL Administrative 

Law Judge’s (ALJ) recommended 

decision (2005–ERA–6) where, 

following an evidentiary hearing, the 

ALJ had concluded that Shaw had not 

violated section 211 of the Energy 

Reorganization Act, as amended, by 

terminating the former painter foreman. 

Shaw denies that it has retaliated 

against the former painter foreman for 

engaging in a protected activity and is 

pursuing its legal challenge to the ARB 

decision. 

In its June 2, 2009 letter, the NRC 

offered Shaw the opportunity to provide 

a written response, attend a predecisional enforcement conference, or 

request ADR in which a neutral 

mediator with no decision-making 

authority would facilitate discussions 

between the NRC and Shaw and, if 

possible, assist the NRC and Shaw in 

reaching an agreement. Shaw requested 

to use ADR to resolve differences it had 

with the NRC. 

On August 24, 2010, the NRC and 

Shaw met in an ADR mediation session 

arranged through the Cornell University 

Institute on Conflict Resolution. This 

Confirmatory Order is issued pursuant 

to the agreement reached during the 

ADR process. 

III 

A. The NRC acknowledged that Shaw, 

for its own business reasons, had 

already put in place during the past 

several years the following policies, 

practices and programs that support 

Safety Conscious Work Environment 

(SCWE) and Safety Culture: 

At the parent company, The Shaw 

Group Inc., level: 

• The SpeakUp Program is a toll-free 

hotline and Web site in which workers 

can report issues to the Company. 

Reports can be made anonymously; 

• The Stop Work Policy gives 

employees authority to immediately 

stop any work activity that presents a 

danger to him/her, co-workers, clients, 

partners or the public without fear of 

reprimand or retaliation; 

• The Targeting Zero Program focuses 

on achieving zero environmental, health 

and safety incidents; it minimizes 

health and safety risks to employees, 

clients, the public and the environment; 

• The Employment Discipline Policy 

prohibits retaliation for exercising the 

right to raise safety concerns; 

• Mandatory Code of Corporate 

Conduct training for all employees with 

computer access; 

• Consideration of integrity and 

compliance as performance factors in 

annual employee performance 

evaluations; 

• Periodic independent culture 

surveys. 

For nuclear maintenance sites: 

• A SCWE Procedure outlines the 

Company’s expectations, and each 

individual’s responsibilities for 

establishing and maintaining a SCWE; 

• The New to Nuclear Workforce 

Orientation Program provides training 

and resources specific to working in the 

nuclear industry for workers coming in 

without nuclear experience; 

• New Hire Orientation informs new 

hires about Shaw’s Safety Culture and 

SCWE expectations, and informs them 

of their responsibilities and programs 

available to them, including SpeakUp 

and Stop Work and Shaw’s nondiscrimination and harassment policies; 

• Supervisor Challenge (Oral Boards) 

evaluates supervisors’ skills in the key 

focus areas including leadership, human 

performance, work performance, and 

reinforcing expectations. 

For new nuclear construction sites: 

• An on-site Employee Concerns 

Program, modeled on resources in NEI 

97–05 is available to all site workers; 

• Procedure Maintaining a Strong 

Nuclear Safety Culture & Safety 

Conscious Work Environment, modeled 

on NEI 09–12 and RIS 2005–018, 

describes Shaw’s expectations for a 

SCWE and the methods by which it will 

establish and maintain it; 

• Shaw provides SCWE training 

comprised of four modules: Inprocessing for all personnel; 90-day 

enhanced training with case studies for 

new craft personnel; nuclear 

professional for office workers; and 

training for supervisors and above; 

• Shaw conducts periodic SCWE 

surveys based on NEI 09–12 survey tool. 

B. During the ADR mediation session, 

an agreement in principle was reached 

where Shaw agreed to take the following 

additional actions: 

1. Within 2 months of issuance of this 

Confirmatory Order, Shaw will issue a 

written communication from a Shaw Power 

Group senior executive to Shaw employees 

in its Nuclear Services (i.e., construction) and 

Nuclear Maintenance Divisions working at 

nuclear facilities addressing: (a) A recent 

DOL ARB decision that concluded that 

retaliation occurred at a SWCI facility in 

2004; b) that Shaw strives to maintain a 

SCWE; and (c) that nuclear workers have 

multiple avenues in which to raise concerns 

and identifying these avenues. 

2. Where not already required by the 

applicable nuclear facility licensee, Shaw 

will establish an Executive Review Board 

(ERB) that will include management 

personnel at or above the level of the site 

project manager, including legal and/or 

human resources participation, to review all 

proposed significant adverse actions (defined 

as three or more days off without pay up to 

and including termination for cause, but 

excludes reductions-in-force and other 

ordinary layoffs) at any NRC-regulated 

maintenance site to ensure these actions 


srobinson on DSKHWCL6B1PROD with NOTICES