Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 27825 

workforce development system will 

quickly boost its understanding of these 

occupations, significantly increasing the 

number of customers requesting training 

in these areas. 

FOR FURTHER INFORMATION CONTACT: 

Michael Harding, Room 4510–C 

Employment and Training 

Administration, 200 Constitution 

Avenue, NW., Washington, DC 20210. 

Telephone number: 202–693–2921 (this 

is not a toll-free number). Fax: 202–693– 

3015. E-mail: Harding.Michael@dol.gov 

Signed at Washington, DC, this 6th day of 

May 2010. 

Jane Oates, 

Assistant Secretary, Employment and 

Training Administration. 

[FR Doc. 2010–11802 Filed 5–17–10; 8:45 am] 

BILLING CODE 4510–FN–P 

NATIONAL FOUNDATION ON THE 

ARTS AND THE HUMANITIES 

National Endowment for the Arts 

Arts Advisory Panel 

Pursuant to Section 10(a)(2) of the 

Federal Advisory Committee Act (Pub. 

L. 92–463), as amended, notice is hereby 

given that one meeting of the Arts 

Advisory Panel to the National Council 

on the Arts will be held at the Nancy 

Hanks Center, 1100 Pennsylvania 

Avenue, NW., Washington, DC 20506 as 

follows (ending times are approximate): 

Design/Mayor’s Institute on City Design 25th 

Anniversary Initiative 

(Application review): June 3–4, 2010 in 

Room 714. A portion of this meeting, from 

3:30 p.m. to 4:30 p.m. on June 4th, will be 

open to the public for policy discussion. The 

remainder of the meeting, from 9 a.m. to 5:30 

p.m. on June 3rd and from 9 a.m. to 3:30 p.m. 

and from 4:30 p.m. to 5:30 p.m. on June 4th, 

will be closed. 

The closed portions of meetings are 

for the purpose of Panel review, 

discussion, evaluation, and 

recommendations on financial 

assistance under the National 

Foundation on the Arts and the 

Humanities Act of 1965, as amended, 

including information given in 

confidence to the agency. In accordance 

with the determination of the Chairman 

of November 10, 2009, these sessions 

will be closed to the public pursuant to 

subsection (c)(6) of section 552b of Title 

5, United States Code. 

Any person may observe meetings, or 

portions thereof, of advisory panels that 

are open to the public, and if time 

allows, may be permitted to participate 

in the panel’s discussions at the 

discretion of the panel chairman. If you 

need any accommodations due to a 

disability, please contact the Office of 

AccessAbility, National Endowment for 

the Arts, 1100 Pennsylvania Avenue, 

NW., Washington, DC 20506, 202/682– 

5532, TDY–TDD 202/682–5496, at least 

seven (7) days prior to the meeting. 

Further information with reference to 

these meetings can be obtained from Ms. 

Kathy Plowitz-Worden, Office of 

Guidelines & Panel Operations, National 

Endowment for the Arts, Washington, 

DC 20506, or call 202/682–5691. 

Dated: May 13, 2010. 

Kathy Plowitz-Worden, 

Panel Coordinator, Panel Operations, 

National Endowment for the Arts. 

[FR Doc. 2010–11812 Filed 5–17–10; 8:45 am] 

BILLING CODE 7537–01–P 

NATIONAL SCIENCE FOUNDATION 

Proposal Review; Notice of Meetings 

In accordance with the Federal 

Advisory Committee Act (Pub. L. 92– 

463, as amended), the National Science 

Foundation (NSF) announces its intent 

to hold proposal review meetings 

throughout the year. The purpose of 

these meetings is to provide advice and 

recommendations concerning proposals 

submitted to the NSF for financial 

support. The agenda for each of these 

meetings is to review and evaluate 

proposals as part of the selection 

process for awards. The review and 

evaluation may also include assessment 

of the progress of awarded proposals. 

The majority of these meetings will take 

place at NSF, 4201 Wilson Blvd., 

Arlington, Virginia 22230. 

These meetings will be closed to the 

public. The proposals being reviewed 

include information of a proprietary or 

confidential nature, including technical 

information; financial data, such as 

salaries; and personal information 

concerning individuals associated with 

the proposals. These matters are exempt 

under 5 U.S.C. 552b(c), (4) and (6) of the 

Government in the Sunshine Act. NSF 

will continue to review the agenda and 

merits of each meeting for overall 

compliance with the Federal Advisory 

Committee Act. 

These closed proposal review 

meetings will not be announced on an 

individual basis in the Federal Register. 

NSF intends to publish a notice similar 

to this on a quarterly basis. For an 

advance listing of the closed proposal 

review meetings that include the names 

of the proposal review panel and the 

time, date, place, and any information 

on changes, corrections, or 

cancellations, please visit the NSF Web 

site: http://www.nsf.gov. This 

information may also be requested by 

telephoning, 703/292–8182. 

Dated: May 13, 2010. 

Susanne Bolton, 

Committee Management Officer. 

[FR Doc. 2010–11824 Filed 5–17–10; 8:45 am] 

BILLING CODE 7555–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0179] 

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 that 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 April 22 to 

May 5, 2010. The last biweekly notice 

was published on May 4, 2010 (75 FR 

23808). 

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) 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. 

mstockstill on DSKH9S0YB1PROD with NOTICES

27826 Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example, 

in derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

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 basis 

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 

mstockstill on DSKH9S0YB1PROD with NOTICES

Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 27827 

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/applycertificates.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/esubmittals.html. Participants may 

attempt to use other software not listed 

on the Web site, but should note that the 

NRC’s E–Filing system does not support 

unlisted software, and the NRC Meta 

System Help Desk will not be able to 

offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E–Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through EIE, users will be 

required to install a Web browser plugin 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/e-submittals

.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. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois, Docket Nos. STN 

50–454 and STN 50–455, Byron Station, 

Unit Nos. 1 and 2, Ogle County, Illinois 

Date of amendment request: March 

29, 2010. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specification (TS) 

5.5.7, ‘‘Reactor Coolant Pump Flywheel 

Inspection Program,’’ by extending the 

reactor coolant pump (RCP) motor 

flywheel inspection interval for certain 

RCP motors from the currentlyapproved 10-year inspection interval to 

an interval not to exceed 20 years. The 

availability of this TS revision was 

announced in the Federal Register on 

October 22, 2003 (68 FR 60422) as part 

of the consolidated line item 

improvement process. In its application, 

the licensee affirmed the applicability of 

the model no significant hazards 

consideration determination, as 

published in the Federal Register on 

June 24, 2003 (68 FR 37590). 

Basis for proposed no significant 

hazards consideration determination: 

mstockstill on DSKH9S0YB1PROD with NOTICES

27828 Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration adopted by the 

licensee is presented below: 

Criterion 1—The Proposed Change Does 

Not Involve a Significant Increase in the 

Probability or Consequences of an 

Accident Previously Evaluated 

The proposed change to the RCP 

flywheel examination frequency does 

not change the response of the plant to 

any accidents. The RCP will remain 

highly reliable and the proposed change 

will not result in a significant increase 

in the risk of plant operation. Given the 

extremely low failure probabilities for 

the RCP motor flywheel during normal 

and accident conditions, the extremely 

low probability of a loss-of-coolant 

accident (LOCA) with loss of offsite 

power (LOOP), and assuming a 

conditional core damage probability 

(CCDP) of 1.0 (complete failure of safety 

systems), the core damage frequency 

(CDF) and change in risk would still not 

exceed the NRC’s [Nuclear Regulatory 

Commission’s] acceptance guidelines 

contained in RG 1.174 [Regulatory 

Guide 1.174, ‘‘An Approach for Using 

Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific 

Changes to the Licensing Basis’’] (<1.0E– 

6 per year). Moreover, considering the 

uncertainties involved in this 

evaluation, the risk associated with the 

postulated failure of an RCP motor 

flywheel is significantly low. Even if all 

four RCP motor flywheels are 

considered in the bounding plant 

configuration case, the risk is still 

acceptably low. 

The proposed change does not 

adversely affect accident initiators or 

precursors, nor alter the design 

assumptions, conditions, or 

configuration of the facility, or the 

manner in which the plant is operated 

and maintained; alter or prevent the 

ability of structures, systems, 

components (SSCs) from performing 

their intended function to mitigate the 

consequences of an initiating event 

within the assumed acceptance limits; 

or affect the source term, containment 

isolation, or radiological release 

assumptions used in evaluating the 

radiological consequences of an 

accident previously evaluated. Further, 

the proposed change does not increase 

the type or amount of radioactive 

effluent that may be released offsite, nor 

significantly increase individual or 

cumulative occupational/public 

radiation exposure. The proposed 

change is consistent with the safety 

analysis assumptions and resultant 

consequences. Therefore, the proposed 

change does not involve a significant 

increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident From Any 

Accident Previously Evaluated 

The proposed change in flywheel 

inspection frequency does not involve 

any change in the design or operation of 

the RCP. Nor does the change to 

examination frequency affect any 

existing accident scenarios, or create 

any new or different accident scenarios. 

Further, the change does not involve a 

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

new or different type of equipment will 

be installed) or alter the methods 

governing normal plant operation. In 

addition, the change does not impose 

any new or different requirements or 

eliminate any existing requirements, 

and does not alter any assumptions 

made in the safety analysis. The 

proposed change is consistent with the 

safety analysis assumptions and current 

plant operating practice. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3—The Proposed Change Does 

Not Involve a Significant Reduction in 

a Margin of Safety 

The proposed change does not alter 

the manner in which safety limits, 

limiting safety system settings, or 

limiting conditions for operation are 

determined. The safety analysis 

acceptance criteria are not impacted by 

this change. The proposed change will 

not result in plant operation in a 

configuration outside of the design 

basis. The calculated impact on risk is 

insignificant and meets the acceptance 

criteria contained in RG 1.174. There are 

no significant mechanisms for inservice 

degradation of the RCP flywheel. 

Therefore, the proposed change does not 

involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

analysis adopted by the licensee and, 

based on this review, it appears that the 

three standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendments involve no significant 

hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Stephen J. 

Campbell. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of amendment request: March 

19, 2010. 

Description of amendment request: 

This submittal requests changes to 

extend the Technical Specification (TS) 

allowed outage time (AOT) for the Unit 

1 and Unit 2 Suppression Pool Cooling 

(SPC) mode of the Residual Heat 

Removal (RHR) system, the Residual 

Heat Removal Service Water (RHRSW) 

system, the Emergency Service Water 

(ESW) system, and the A.C. SourcesOperating (Emergency Diesel 

Generators) from 72 hours to seven (7) 

days in order to allow for repairs of the 

RHRSW system piping. Specifically, the 

proposal adds a footnote to the affected 

TS limiting conditions for operation to 

indicate that the 72-hour AOT for the 

affected system may be extended once 

per calendar year, for one unit only, for 

a period of up to 7 days to allow for 

repairs of one RHRSW subsystem piping 

with the opposite unit shutdown, 

reactor vessel head removed and reactor 

cavity flooded, and other specific 

compensatory measures in effect. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee (Exelon) has provided its 

analysis of the issue of no significant 

hazards consideration, which is 

presented below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed TS changes will not increase 

the probability of an accident since they will 

only extend the time period that one RHRSW 

subsystem, one loop of SPC, one ESW loop 

and two Emergency Diesel Generators (EDGs) 

can be out of service. The extension of the 

time duration that one RHRSW, one ESW 

loop and two EDGs are out of service has no 

direct physical impact on the plant. The 

proposed inoperable RHRSW subsystem, 

ESW loop and two EDGs are normally in a 

standby mode while the unit is in 

[Operational Condition] OPCON 1 or 2 and 

are not directly supporting plant operation. 

Therefore, they can have no impact on the 

plant that would make an accident more 

likely to occur due to their inoperability. 

During transients or events which require 

these subsystems to be operating, there is 

sufficient capacity in the operable loops/ 

subsystems and available[,] but inoperable[,] 

equipment to support plant operation or 

shutdown. Therefore, failures that are 

accident initiators will not occur more 

frequently than previously postulated as a 

result of the proposed changes. 

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Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 27829 

In addition, the consequences of an 

accident previously evaluated in the Updated 

Final Safety Analysis Report (UFSAR) will 

not be increased. With one RHRSW 

subsystem inoperable, one SPC loop, one 

ESW loop and two EDGs inoperable but 

verified available prior to entering the 

proposed configuration, a known quantity of 

equipment is inoperable. Based on the 

support functions of the RHRSW system, a 

review of the plant was performed to 

determine the impacts that the inoperable 

RHRSW subsystem would have on other 

systems. The impacts were identified for 

each system and it was determined whether 

there were any adverse effects on the 

systems. It was then determined how the 

adverse effects would impact each system’s 

design basis and overall plant safety. The 

consequences of any postulated accidents 

occurring on Unit 1 or Unit 2 during these 

AOT extensions was found to be bounded by 

the previous analyses as described in the 

UFSAR. Since the inoperable ESW loop, 

selected emergency core cooling system 

(ECCS) pumps and EDGs will be verified 

available prior to entering the proposed 

configuration, they would have no impact on 

other systems. 

The minimum equipment required to 

mitigate the consequences of an accident 

and/or safely shut down the plant will be 

operable or available. Therefore, by 

extending certain AOTs and extending the 

assumptions concerning the combinations of 

events for the longer duration of each 

extended AOT, Exelon concludes that at least 

the minimum equipment required to mitigate 

the consequences of an accident and/or 

safely shut down the plant will still be 

operable or available during the extended 

AOT. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed TS changes will not create 

the possibility of a different type of accident 

since they will only extend the time period 

that one RHRSW subsystem and one loop of 

SPC can be out of service, and one ESW loop 

and two EDGs can be inoperable, but verified 

available, prior to entering the proposed 

configuration. The extension of the time 

duration that one RHRSW subsystem and one 

SPC loop is out of service, and one ESW loop 

and two EDGs are inoperable, but verified 

available, prior to entering the proposed 

configuration has no direct physical impact 

on the plant and does not create any new 

accident initiators. The systems involved are 

accident mitigation systems. All of the 

possible impacts that the inoperable 

equipment may have on its supported 

systems were previously analyzed in the 

UFSAR and are the basis for the present TS 

Action statements and AOTs. The impact of 

inoperable support systems for a given time 

duration was previously evaluated and any 

accident initiators created by the inoperable 

systems was evaluated. The lengthening of 

the time duration does not create any 

additional accident initiators for the plant. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The present RHRSW, SPC, ESW and EDG 

AOT limits were set to ensure that sufficient 

safety-related equipment is available for 

response to all accident conditions and that 

sufficient decay heat removal capability is 

available for a loss of coolant accident 

(LOCA) coincident with a loss of offsite 

power (LOOP) on one unit and simultaneous 

safe shutdown of the other unit. A slight 

reduction in the margin of safety is incurred 

during the proposed extended AOT due to 

the increased risk that an event could occur 

in a 7-day period versus a 72-hour period. 

This increased risk is judged to be minimal 

due to the low probability of an event 

occurring during the extended AOT and 

based on the following discussion of 

minimum ECCS/decay heat removal 

requirements. 

The inoperable ESW loop, selected ECCS 

pumps and EDGs will be verified available 

prior to entering the proposed configuration; 

therefore, extension of the AOT will have no 

effect on the minimum ECCS equipment 

available or margin of safety. 

The reduction in the margin of safety from 

the extension of the RHRSW, SPC, ESW and 

EDG AOT limits is not significant since the 

remaining operable ECCS equipment is 

adequate to mitigate the consequences of any 

accident. This conclusion is based on the 

information contained in General Electric 

Company documents NEDO–24708A, 

‘‘Additional Information Required for NRC 

Staff Generic Report on Boiling Water 

Reactors,’’ Revision 1, dated December 1980, 

and NEDC[–]3093P–A, ‘‘BWR Owner’s Group 

Technical Specification Improvement 

Methodology (with Demonstration for BWR 

ECCS Activation Instrumentation),’’ dated 

December 1988. These documents describe 

the minimum requirements to successfully 

terminate a transient or LOCA initiating 

event (with scram), assuming multiple 

failures with realistic conditions, and were 

used to justify certain TS AOTs per UFSAR 

Sections 6.3.1.1.2.o and 6.3.3.1. The 

minimum requirements for short-term 

response to an accident would be either one 

Low Pressure Coolant Injection (LPCI) pump 

or one Core Spray subsystem in conjunction 

with Automatic Depressurization System 

(ADS), or the High Pressure Coolant Injection 

(HPCI) system, which would be adequate to 

re-flood the vessel and maintain core cooling 

sufficient to preclude fuel damage. For longterm response, the minimum requirements 

would be one loop of RHR for decay heat 

removal, along with another low-pressure 

ECCS subsystem. These minimum 

requirements will be met since 

implementation of the proposed TS changes 

will require the operability or availability of 

HPCI, ADS, two LPCI subsystems (or one 

LPCI subsystem and one RHR subsystem 

during decay heat removal) and one Core 

Spray subsystem be maintained during the 7- 

day period. Operations personnel are fully 

qualified by normal periodic training to 

respond to and mitigate a Design Basis 

Accident, including the actions needed to 

ensure decay heat removal while LGS Unit 1 

and Unit 2 are in the operational 

configurations described within this 

submittal. Accordingly, procedures are 

already in place that address safe plant 

shutdown and decay heat removal for 

situations applicable to those in the proposed 

AOTs. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1, Dauphin 

County, Pennsylvania 

Date of amendment request: March 

24, 2010. 

Description of amendment request: 

The proposed amendment would 

modify the Three Mile Island, Unit 1 

(TMI–1) Technical Specifications (TSs) 

by relocating specific surveillance 

frequencies to a new licensee-controlled 

program called the Surveillance 

Frequency Control Program. This 

change incorporates the adoption of 

Nuclear Energy Institute (NEI) 04–10, 

‘‘Risk-Informed Technical Specifications 

Initiative 5b, Risk-Informed Method for 

Control of Surveillance Frequencies,’’ 

Revision (Rev.) 1. A description of the 

Surveillance Frequency Control 

Program will be added to the TMI–1 

TSs. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed changes relocate the 

specified frequencies for periodic 

surveillance requirements to licensee control 

under a new Surveillance Frequency Control 

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Program [SFCP]. Surveillance frequencies are 

not an initiator to any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. The systems and 

components required by the technical 

specifications for which the surveillance 

frequencies are relocated are still required to 

be operable, meet the acceptance criteria for 

the surveillance requirements, and be 

capable of performing any mitigation 

function assumed in the accident analysis. 

As a result, the consequences of any accident 

previously evaluated are not significantly 

increased. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

No new or different accidents result from 

utilizing the proposed changes. The changes 

do not involve a physical alteration of the 

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

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements. 

The changes do not alter assumptions made 

in the safety analysis. The proposed changes 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

Response: No. 

The design, operation, testing methods, 

and acceptance criteria for systems, 

structures, and components (SSCs), specified 

in applicable codes and standards (or 

alternatives approved for use by the [Nuclear 

Regulatory Commission] NRC) will continue 

to be met as described in the plant licensing 

basis (including the final safety analysis 

report and bases to TS), since these are not 

affected by changes to the surveillance 

frequencies. Similarly, there is no impact to 

safety analysis acceptance criteria as 

described in the plant licensing basis. To 

evaluate a change in the relocated 

surveillance frequency, Exelon will perform 

a probabilistic risk evaluation using the 

guidance contained in NRC approved NEI 

04–10, Rev. 1, in accordance with the TS 

SFCP. NEI 04–10, Rev. 1, methodology 

provides reasonable acceptance guidelines 

and methods for evaluating the risk increase 

of proposed changes to surveillance 

frequencies consistent with Regulatory Guide 

1.177. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

Exelon Generation Company, LLC, and 

PSEG Nuclear, LLC, Docket No. 50–277, 

Peach Bottom Atomic Power Station 

(PBAPS), Unit 2, York and Lancaster 

Counties, Pennsylvania 

Date of amendment request: August 

28, 2009, as supplemented by letter 

dated February 25, 2010. 

Description of amendment request: 

The proposed change would modify the 

PBAPS Unit 2 Technical Specification 

(TS) Section 5.5.12 to reflect a one-time 

extension of the Type A containment 

Integrated Leak Rate Test (ILRT) to no 

later than October 2015. The proposed 

TS revision would allow a one-time 

extension of 5 years to the 10-year 

frequency of the performance-based 

leakage rate testing program for the 

PBAPS Unit 2 containment Type A 

ILRT test. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change involves a one-time 

extension of the Primary Containment ILRT 

interval from 10 years to 15 years. The 

proposed change does not involve a physical 

change to the plant [* * *]. The Primary 

Containment function is to provide an 

essentially leak tight barrier against the 

uncontrolled release of radioactivity to the 

environment for postulated accidents. As 

such, the containment itself and the testing 

requirements to periodically demonstrate the 

integrity of the containment exist to ensure 

the plant’s ability to mitigate the 

consequences of an accident, and do not 

involve any accident precursors or initiators. 

Therefore, the probability of occurrence of an 

accident previously evaluated is not 

significantly increased by the proposed 

change. 

Continued containment integrity is assured 

by the established programs for local leak 

rate testing and inservice/containment 

inspections, which are unaffected by the 

proposed change. As documented in 

NUREG–1493, ‘‘Performance-Based 

Containment Leak-Test Program,’’ dated 

September 1995, industry experience has 

shown that local leak rate tests (Type B and 

C) have identified the vast majority of 

containment leakage paths, and that ILRTs 

detect only a small fraction of containment 

leakage pathways. 

The potential consequences of the 

proposed change have been quantified by 

analyzing the changes in risk that would 

result from extending the ILRT interval from 

10 years to 15 years. Increasing the ILRT 

interval to 15 years for this one-time change 

is considered to be insignificant since it 

represents a very small change to the PBAPS, 

Unit 2 risk profile. Additionally, the 

proposed change maintains defense-in-depth 

by preserving a reasonable balance among 

prevention of core damage, prevention of 

containment failure, and consequence 

mitigation. PBAPS, Unit 2 has determined 

that the increase in conditional containment 

failure probability due to the proposed 

change is very small. Therefore, it is 

concluded that the proposed one-time 

extension of the Primary Containment ILRT 

interval from 10 years to 15 years does not 

significantly increase the consequences of an 

accident previously evaluated. 

Based on the above discussion, it is 

concluded that 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 involves a one-time 

extension of the Primary Containment ILRT 

interval. The containment and the testing 

requirements to periodically demonstrate the 

integrity of the containment exist to ensure 

the plant’s ability to mitigate the 

consequences of an accident, and do not 

involve any accident precursors or initiators. 

The proposed change does not involve a 

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

different type of equipment will be 

installed)[* * *]. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed one-time extension of the 

Primary Containment ILRT interval does not 

alter the manner in which safety limits, 

limiting safety system setpoints, or limiting 

conditions for operation are determined. The 

specific requirements and conditions of the 

10 CFR 50 Appendix J testing program plan, 

as defined in the Technical Specifications, 

exist to ensure that the degree of Primary 

Containment structural integrity and leaktightness that is considered in the plant 

safety analyses is maintained. The overall 

containment leakage rate limit specified by 

the Technical Specifications is maintained, 

and Type B and C containment leakage tests 

will continue to be performed at the 

frequency currently required by the TS. 

Containment inspections performed in 

accordance with [the * * *] plant programs 

[described above] serve to provide a high 

degree of assurance that the containment will 

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not degrade in a manner that is detectable 

only by an ILRT. Furthermore, a risk 

assessment using the current PBAPS, Unit 2 

Probabilistic Risk Assessment internal events 

model concluded that extending the ILRT 

test interval from 10 years to 15 years results 

in a very small change to the PBAPS, Unit 

2 risk profile. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review with the NRC staff changes noted 

in square brackets above, 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: Mr. J. Bradley 

Fewell, Associate General Counsel, 

Exelon Generation Company LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

Florida Power and Light Company 

(FPL), Docket Nos. 50–250 and 50–251, 

Turkey Point Plant, Units 3 and 4, 

Miami-Dade County, Florida 

Date of amendment request: February 

16, 2010. 

Description of amendment request: To 

revise the licensing bases by removing 

two technical specifications (TSs) that 

restrict movements of heavy loads over 

the spent fuel pools. 

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: 

TS 3/4.9.7, Crane Travel-Spent Fuel 

Storage Areas (reviewed for both units) 

FPL has evaluated whether or not a 

significant hazards consideration is involved 

with removing the TS 3/4.9.7, ‘‘Crane 

Travel—Spent Fuel Storage Areas,’’ from the 

Turkey Point Units 3 and 4 TS by focusing 

on the three standards set forth in 10 CFR 

50.92, ‘‘Issuance of amendment,’’ as discussed 

below: 

(1) Would operation of the facility in 

accordance with the proposed amendment 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

Response: No. 

The removal of TS 3/4.9.7 will not increase 

the probability of a fuel handling accident 

(FHA), as evaluated in Chapter 14.2.1 of the 

UFSAR [Updated Final Safety Analysis 

Report], and is considered remote because of 

the administrative controls and physical 

limitations imposed on fuel handling 

operations. The load limit restriction, in 

conjunction with existing plant documents 

(for example, Turkey Point heavy load 

handling procedures) that restrict crane or 

other heavy load handling operations provide 

a defense-in-depth approach to handling 

heavy loads in the spent fuel pool vicinity. 

The load limitation defined in TS 3/4.9.7 is 

preserved and will be implemented based on 

the operation limits and safety margins for 

the control of heavy loads consistent with 

NUREG–0612. The TS change does not 

represent any physical change to the plant 

systems, structures, or components. 

Therefore, the systems credited with 

mitigating the dose consequences of a FHA 

remain in place. The dose consequences of a 

fuel handling accident as discussed in 

Turkey Point UFSAR Chapter 14.2.1 will not 

increase because of the administrative 

controls and physical limitations imposed on 

fuel handling operations which minimize the 

likelihood of a FHA. 

Therefore, facility operation in accordance 

with the proposed amendment would not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

(2) Would operation of the facility in 

accordance with the proposed amendment 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated? 

Response: No. 

The removal of TS 3/4.9.7 does not 

represent any physical change to the plant 

systems, structures, or components. The 

same operational functions of moving new 

fuel, spent fuel, or other loads over the spent 

fuel pool are retained and therefore do not 

create or increase the possibility of a new or 

different kind of accident from any accident 

previously evaluated. Additionally, the load 

limit of 2000 pounds over the spent fuel pool 

defined in TS 3/4.9.7 is preserved and 

implemented in existing plant documents 

and are established based on the operational 

limits and safety margins for the control of 

heavy loads consistent with NUREG–0612. 

Other measures which preclude the creation 

of a new or different type of accident include 

interlocks and physical stops, operator 

training, and load handling procedures. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

(3) Would operation of the facility in 

accordance with the proposed amendment 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The removal of TS 3/4.9.7 does not change 

the operational process of moving loads over 

the spent fuel pool. There are no changes to 

any physical plant systems, structures, or 

components. The spent fuel handling crane 

has weight sensors that are interlocked to 

limit the total load. In addition, an in-line 

weight sensing system is provided for each 

hoist to limit the lifting load to preclude 

accidental fuel damage should binding occur. 

When lifting over spent fuel, the total load 

is limited to 2000 pounds by current 

procedures, limit switches and load sensors. 

Because of these measures, no margin of 

safety is reduced or compromised. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

will not involve a significant reduction in a 

margin of safety. 

Based on the above, FPL concludes 

that the proposed amendment does not 

involve a significant hazards 

consideration under the standards set 

forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

TS 3/4.9.12, Handling of Spent Fuel 

Cask (reviewed for both units) 

FPL has evaluated whether or not a 

significant hazards consideration is 

involved with the proposed amendment 

of removing TS 3/4.9.12, ‘‘Handling of 

Spent Fuel Cask,’’ by focusing on the 

three standards set forth in 10 CFR 

50.92, ‘‘Issuance of amendment,’’ as 

discussed below: 

(1) Would operation of the facility in 

accordance with the proposed amendment 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

Response: No. 

The removal of TS 3/4.9.12 will not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. The accident evaluated 

for the existing spent fuel cask handling 

crane is the drop of a single element cask as 

cited in UFSAR Section 14.2.1.3, ‘‘Cask Drop 

Accident.’’ This cask drop accident was 

analyzed and the radiological dose 

consequence, as a result of the cask drop, is 

determined to be within the limits of 10 CFR 

100. The current spent fuel cask handling 

crane at Turkey Point Units 3 and 4 has a 

single 105/15 ton main/auxiliary hook design 

capacity and is not designed as single-failureproof. The new spent fuel cask handling 

crane will be single-failure-proof meeting all 

of the requirements of NUREG–0554, ‘‘Single 

Failure Proof Cranes for Nuclear Power 

Plants’’ and also NUREG–0612, Section 5.1.6, 

‘‘Single Failure Proof Handling Systems.’’ The 

probability of a cask drop accident using a 

single-failure-proof crane designed and 

operated to these NUREG requirements is 

considered to be extremely small. 

The design for the upgrade of the spent 

fuel cask handling crane is to increase the 

capacity to 130/25 tons (main/auxiliary 

hook). All crane components (hoist, bridge, 

girders, etc.) are designed and fabricated to 

retain control of and hold the maximum 

critical load (a planned 32 element spent fuel 

cask) in the unlikely event of the failure of 

a single component, coincident with a Design 

or Maximum earthquake. 

The objectives cited in Section 5.1 of 

NUREG–0612, ‘‘Recommended Guidelines,’’ 

for the control of heavy loads are satisfied. 

The probability of a cask drop accident using 

the new single-failure-proof spent fuel cask 

crane, as compared to the existing nonsingle-failure-proof crane, is therefore not 

increased. The increase of the consequences 

of an accident previously evaluated is also 

not increased because the potential for a cask 

drop by the new upgraded spent fuel cask 

handling crane is considered to be extremely 

small. 

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Further, operational limits, interlocks, 

procedural and administrative controls, that 

restrict the handling of heavy loads over fuel 

stored in the spent fuel pool, provide 

additional defense-in depth to ensure that a 

load could not be dropped that would result 

in dose consequences greater than previously 

evaluated. 

It is concluded that facility operation in 

accordance with the proposed amendment 

would not involve a significant increase in 

the probability or consequences of an 

accident previously evaluated. 

(2) Would operation of the facility in 

accordance with the proposed amendment 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated? 

Response: No. 

Operation of the spent fuel cask handling 

crane after the upgrade to a single-failureproof design will remain the same as the 

operation of the existing spent fuel cask 

handling crane. The distinction is the load 

that will be lifted. 

The new spent fuel cask is a multiple 

assembly cask, in contrast to a single 

assembly cask as currently specified for use. 

The current spent fuel cask handling crane is 

designed to lift a single element spent fuel 

cask. The upgraded capacity of the new spent 

fuel cask handling crane will allow for lifting 

a cask designed to hold a maximum of 32 

spent fuel assemblies. Current operating and 

administrative procedures that restrict the 

movement of heavy loads over fuel stored in 

the spent fuel pool remain in place. The new 

spent fuel cask handling crane is designed, 

fabricated and tested to single-failure-proof 

requirements (NUREG–0554, ‘‘Single Failure 

Proof Cranes for Nuclear Power Plants’’ and 

NUREG–0612, Section 5.1.6, ‘‘Single Failure 

Proof Handling Systems’’) and will be 

operated within the procedural and 

administrative framework as the currently 

installed spent fuel cask handling crane. 

Therefore, the possibility of a new or 

different kind of accident from any accident 

previously evaluated is not created from the 

removal of TS 3/4.9.12. 

Therefore, it can be concluded that the 

operation of the facility in accordance with 

the proposed amendment would not create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

(3) Would operation of the facility in 

accordance with the proposed amendment 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The existing spent fuel cask handling crane 

is not designed as single-failure-proof in 

accordance with NUREG–0612. The new 

spent fuel cask handling crane is designed, 

and will be fabricated, installed and tested to 

the single-failure-proof requirements as 

outlined in NUREG–0612, Section 5.1.6, 

‘‘Single Failure Proof Handling Systems.’’ The 

use of the defense-in-depth approach for the 

control and handling of heavy loads as cited 

in Section 5.1 of NUREG–0612, 

‘‘Recommended Guidelines,’’ provides 

assurance that there is a sufficient margin of 

safety in the handling of heavy loads. 

Thereby, the removal of TS 3/4.9.12 will not 

involve a significant reduction in the margin 

of safety. 

Defense-in-depth measures include 

operational limits, interlocks, procedural and 

administrative controls, rigging, load paths, 

testing, training, maintenance and other 

related considerations. These measures 

provide assurance that the margin of safety 

is not reduced in the operation of the facility 

by meeting all the requirements of NUREG– 

0612 and NUREG–0554. The specific 

requirements and FPL compliance with them 

is documented in the NUREG–0554 

Compliance Matrix [Attachment 3 to this 

application]. 

The design for the upgrade of the spent 

fuel cask handling crane is to increase the 

capacity to 130/25 tons (main/auxiliary 

hook). The spent fuel cask handling crane 

has a Main Hoist and Auxiliary Hoist Cable 

Safety Factor of a minimum 10:1 on nominal 

breaking strength at 130 tons and 25 tons 

respectively and is fully compliant with 

ASME NOG–1 Section 5425.1. The Main 

Hoist Hook and Auxiliary Hoist Hook Safety 

Factor have a 10:1 minimum on ultimate 

strength at 130 tons and 25 tons, respectively. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

will not involve a significant reduction in a 

margin of safety. 

Based on the above, FPL concludes 

that the proposed amendment does not 

involve a significant hazards 

consideration under the standards set 

forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: M.S. Ross, 

Attorney, Florida Power & Light, P.O. 

Box 14000, Juno Beach, Florida 33408– 

0420. 

NRC Acting Branch Chief: Douglas A. 

Broaddus. 

[Southern California Edison Company, et al., 

Docket Nos. 50–361 and 50–362, San Onofre 

Nuclear Generating Station, Units 2 and 3, 

San Diego County, California 

Date of amendment request: January 

14, 2010. 

Description of amendment request: 

The amendments would revise a 

number of Technical Specification (TS) 

requirements, to impose similar 

restrictions on the movement of nonirradiated fuel assemblies to those 

currently in place for movement of 

irradiated fuel assemblies. The 

additional restrictions will limit the 

movement of all fuel assemblies over 

irradiated fuel assemblies in 

containment or in the fuel storage pool. 

The affected TS Limiting Conditions for 

Operation (LCOs) are: LCO 3.3.8, 

‘‘Containment Purge Isolation Signal 

(CPIS),’’ LCO 3.3.9, ‘‘Control Room 

Isolation Signal (CRIS),’’ LCO 3.7.11, 

‘‘Control Room Emergency Air Cleanup 

System (CREACUS),’’ LCO 3.7.16, ‘‘Fuel 

Storage Pool Water Level,’’ LCO 3.8.2, 

‘‘AC Sources—Shutdown,’’ LCO 3.8.5, 

‘‘DC Sources—Shutdown,’’ LCO 3.8.8, 

‘‘Inverters—Shutdown,’’ LCO 3.8.10, 

‘‘Distribution Systems—Shutdown,’’ 

LCO 3.9.3, ‘‘Containment Penetrations,’’ 

and LCO 3.9.6, ‘‘Refueling Water Level.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This proposed change revises Technical 

Specifications applicability wording 

regarding the movement of fuel assemblies in 

containment and the fuel storage pool at the 

San Onofre Nuclear Generating Station 

(SONGS) Units 2 and 3 to include the 

movement of both irradiated and nonirradiated fuel assemblies. The proposed 

applicability is more comprehensive than the 

current Applicability. 

Expanding the applicability of the relevant 

Technical Specifications is necessary to 

account for updated fuel drop analyses 

which demonstrate that impacted spent fuel 

assemblies may be damaged. Consequently, 

movement of non-irradiated fuel assemblies 

could result in a Fuel Handling Accident that 

has radiological consequences. Changing the 

applicability of the relevant Technical 

Specifications does not affect the probability 

of a Fuel Handling Accident. The expanded 

applicability provides assurance that 

equipment designed to mitigate a Fuel 

Handling Accident is capable of performing 

its specified safety function, such that the 

consequences of an accident are not 

increased. 

Consequently, 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 revised spent fuel drop analyses 

demonstrate that impacted fuel assemblies 

may be damaged as the result of a dropped 

fuel assembly. The existing SONGS 

Technical Specifications regarding 

movement of fuel assemblies are not 

applicable for movement of non-irradiated 

fuel assemblies. A drop of a non-irradiated 

fuel assembly that has radiological 

consequences could occur during periods 

when equipment that would be required to 

mitigate those consequences is not required 

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to be OPERABLE in accordance with the 

existing Technical Specifications. 

The proposed changes to the Technical 

Specifications applicability language 

regarding the movement of fuel assemblies in 

containment and the fuel storage pool at 

SONGS Units 2 and 3 ensure that Limiting 

Conditions of Operation and appropriate 

Required Actions for required equipment are 

in effect during fuel movement. This 

provides assurance that any Fuel Handling 

Accident that may occur will remain within 

the initial assumptions of accident analyses. 

Consequently, there is no possibility of a 

new or different kind of accident due to this 

change. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed Technical Specifications 

change will not affect protection criterion for 

plant equipment and will not reduce the 

margin of safety. By extending the 

Applicability to the movement of nonirradiated fuel assemblies, the current margin 

of safety is maintained. 

Consequently, there is no significant 

reduction in a margin of safety due to this 

change. 

The NRC staff has reviewed the 

licensee’s analysis and, based on that 

review, it appears that the three 

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: Douglas K. 

Porter, Esquire, Southern California 

Edison Company, 2244 Walnut Grove 

Avenue, Rosemead, California 91770. 

NRC Branch Chief: Michael T. 

Markley. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: 

November 25, 2009. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.3.2, 

‘‘Engineered Safety Feature Actuation 

System (ESFAS) Instrumentation,’’ that 

would add a new Required Action Q.1 

to require restoration of an inoperable 

Balance of Plant (BOP) ESFAS train to 

OPERABLE status within 24 hours. In 

addition, the Completion Times for TS 

3.3.2 Required Actions J.1 and O.1 to 

trip inoperable channels that provide 

inputs to BOP ESFAS would also be 

extended to 24 hours. Shutdown track 

Completion Times to be in MODES 3 

and 4 would be increased to reflect 

longer restoration times. Separate 

Condition entry for TS Condition J 

would be restricted to assure that 

Function 6.g in TS Table 3.3.2–1 will 

provide a start signal to the motordriven auxiliary feedwater pumps from 

one train of BOP ESFAS actuation logic. 

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. 

Overall protection system performance will 

remain within the bounds of the previously 

performed accident analyses since no 

hardware changes are proposed to the 

protection systems. The same reactor trip 

system (RTS) and engineered safety feature 

actuation system (ESFAS) instrumentation 

will continue to be used. The protection 

systems will continue to function in a 

manner consistent with the plant design 

basis. There will be no changes to the BOP 

ESFAS surveillance and operating limits. 

The proposed changes will not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, and 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes will not alter or 

prevent the ability of structures, systems, and 

components (SSCs) from performing their 

intended functions to mitigate the 

consequences of an initiating event within 

the assumed acceptance limits. 

The proposed changes do not affect the 

way in which safety-related systems perform 

their functions. 

All accident analysis acceptance criteria 

will continue to be met with the proposed 

changes. The proposed changes will not 

affect the source term, containment isolation, 

or radiological release assumptions used in 

evaluating the radiological consequences of 

an accident previously evaluated. The 

proposed changes will not alter any 

assumptions or change any mitigation actions 

in the radiological consequence evaluations 

in the FSAR [Final Safety Analysis Report]. 

The applicable radiological dose 

acceptance criteria will continue to be met. 

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. 

There are no proposed changes in the 

method by which any safety-related plant 

SSC performs its safety function. The 

proposed changes will not affect the normal 

method of plant operation or change any 

operating parameters. No equipment 

performance requirements will be affected. 

The proposed changes will not alter any 

assumptions made in the safety analyses. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures will be introduced as a result 

of this amendment. There will be no adverse 

effect or challenges imposed on any safetyrelated system as a result of this amendment. 

The proposed amendment will not alter the 

design or performance of the 7300 Process 

Protection System, Nuclear Instrumentation 

System, Solid State Protection System, BOP 

ESFAS, MSFIS [main steam/feedwater 

isolation system], or LSELS [load shedder 

and emergency load sequencer] used in the 

plant protection systems. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

There will be no effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, heat flux hot channel factor 

(FQ), nuclear enthalpy rise hot channel factor 

(FDH), loss of coolant accident peak cladding 

temperature (LOCA PCT), peak local power 

density, or any other margin of safety. The 

applicable radiological dose consequence 

acceptance criteria will continue to be met. 

The proposed changes do not eliminate 

any surveillances or alter the frequency of 

surveillances required by the Technical 

Specifications. No instrument setpoints or 

system response times are affected. None of 

the acceptance criteria for any accident 

analysis will be changed. 

The proposed changes will have no impact 

on the radiological consequences of a design 

basis accident. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units No. 1 and 

No. 2, Louisa County, Virginia 

Date of amendment request: March 

30, 2010. 

Description of amendment request: 

The proposed amendments would 

modify the North Anna Technical 

Specifications (TSs) 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.’’ The changes 

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27834 Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 

are consistent with NRC-approved 

Industry Technical Specifications Task 

Force (TSTF) Standard Technical 

Specifications (STS) change TSTF–425, 

Revision 3. The Federal Register notice 

published on July 6, 2009 (74 FR 

31996), announced the availability of 

this TS improvement. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed changes relocate the 

specified frequencies for periodic 

surveillance requirements to licensee control 

under a new Surveillance Frequency Control 

Program. Surveillance frequencies are not an 

initiator to any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. The systems and 

components required by the technical 

specifications for which the surveillance 

frequencies are relocated are still required to 

be operable, meet the acceptance criteria for 

the surveillance requirements, and be 

capable of performing any mitigation 

function assumed in the accident analysis. 

As a result, the consequences of any accident 

previously evaluated are not significantly 

increased. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

No new or different accidents result from 

utilizing the proposed changes. The changes 

do not involve a physical alteration of the 

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

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements. 

The changes do not alter assumptions made 

in the safety analysis. The proposed changes 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

Response: No. 

The design, operation, testing methods, 

and acceptance criteria for systems, 

structures, and components (SSCs), specified 

in applicable codes and standards (or 

alternatives approved for use by the NRC) 

will continue to be met as described in the 

plant licensing basis (including the final 

safety analysis report and bases to TS), since 

these are not affected by changes to the 

surveillance frequencies. Similarly, there is 

no impact to safety analysis acceptance 

criteria as described in the plant licensing 

basis. To evaluate a change in the relocated 

surveillance frequency, Dominion will 

perform a probabilistic risk evaluation using 

the guidance contained in NRC approved NEI 

04–10, Rev. 1 in accordance with the TS 

SFCP. NEI 04–10, Rev. 1, methodology 

provides reasonable acceptance guidelines 

and methods for evaluating the risk increase 

of proposed changes to surveillance 

frequencies consistent with Regulatory Guide 

1.177. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., 120 Tredegar 

Street, RS–2, Richmond, VA 23219. 

NRC Branch Chief: Gloria Kulesa. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices, either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

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: 

June 25, 2008, as supplemented on 

November 6, 2008, March 9, 2009, June 

12, 2009, December 18, 2009, and March 

26, 2010. 

Brief description of amendment 

request: The proposed amendment 

would revise the PBAPS, Units 2 and 3, 

Technical Specification Section 

4.3.1.1.a concerning the spent fuel pool 

k-infinity value. 

Date of publication of individual 

notice in Federal Register: April 26, 

2010 (75 FR 21680). 

Expiration date of individual notice: 

May 26, 2010 (comment request); June 

25, 2010 (hearing request). 

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 

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Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 27835 

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. 

Dominion Nuclear Connecticut, Inc., et 

al., Docket No. 50–423, Millstone Power 

Station, Unit No. 3, New London 

County, Connecticut 

Date of application for amendment: 

November 23, 2009, as supplemented by 

letter dated April 26, 2010. 

Brief description of amendment: The 

license amendment request revises the 

Millstone Power Station, Unit 3 (MPS3) 

Technical Specification (TS) 6.8.4.g, 

‘‘Steam Generator Program,’’ to exclude 

a portion of the tubes below the top of 

the steam generator tubesheet from 

periodic steam generator tube 

inspections. This request also removes 

reference to the previous Cycle 13 

interim alternate repair criteria. 

Date of issuance: May 3, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: 249. 

Renewed Facility Operating License 

No. NPF–49: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: January 26, 2010 (75 FR 

4114). The supplemented dated April 

26, 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 May 3, 2010. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of application for amendments: 

December 1, 2008. 

Brief description of amendments: The 

amendments correct a non-conservative 

Technical Specification (TS) 

Surveillance Requirement by revising 

McGuire TS 3.8.1.4 to increase the 

minimum required amount of fuel oil 

for the Emergency Diesel Generators 

fuel oil day tank as read on the local 

fuel gauge used to perform the 

surveillance. 

Date of issuance: May 5, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 254 and 234. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: May 19, 2009 (74 FR 23442). 

The supplements dated July 30, 2009, 

December 2, 2009, and March 10, 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 May 5, 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 7, 2008, as supplemented on 

May 7, 2009, and January 19, 2010. 

Brief description of amendments: The 

August 7, 2008, submittal contained 

several areas of review that are being 

dispositioned as separate amendment 

requests. The amendments associated 

with this notice revise the PBAPS Units 

2 and 3 Technical Specifications (TS) to 

delete the list of emergency diesel 

generator critical trips from TS 

Surveillance Requirement (SR) 3.8.1.13 

and clarify that the purpose of the SR is 

to verify that the non-critical trips are 

bypassed. This TS change adopts 

Technical Specification Task Force 

(TSTF) Traveler 400, Revision 1, 

‘‘Clarify SR on Bypass of DG [diesel 

generator] Automatic Trips.’’ 

Date of issuance: April 30, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 275 and 279. 

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, 2009 (74 FR 20744). 

The supplements dated May 7, 2009, 

and January 19, 2010, clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the initial proposed 

no significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 30, 2010. 

No significant hazards consideration 

comments received: No. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power Plant, 

Unit Nos. 1 and 2, Somervell County, 

Texas 

Date of amendment request: April 2, 

2009. 

Brief description of amendments: The 

amendment revised Technical 

Specification (TS) 3.3.1 entitled, 

‘‘Reactor Trip System (RTS) 

Instrumentation’’ to add Surveillance 

Requirement 3.3.1.16 to Function 3 of 

TS Table 3.3.1–1 to verify that the RTS 

response times are within limits every 

18 months on staggered basis. The 

change is based on a reanalysis of the 

Rod Cluster Control Assembly Bank 

Withdrawal at Power event. 

Date of issuance: April 26, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: Unit 1—151; Unit 

2–151. 

Facility Operating License Nos. NPF– 

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: May 19, 2009 (74 FR 23446). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 26, 2010. 

No significant hazards consideration 

comments received: No. 

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and 

Opportunity for a Hearing (Exigent 

Public Announcement or Emergency 

Circumstances) 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application for the 

amendment complies with the 

standards and requirements of the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s rules 

and regulations. The Commission has 

made appropriate findings as required 

by the Act and the Commission’s rules 

and regulations in 10 CFR Chapter I, 

which are set forth in the license 

amendment. 

Because of exigent or emergency 

circumstances associated with the date 

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27836 Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order. 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, 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’s (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. 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. Within 

60 days after the date of publication of 

this notice, any person(s) whose interest 

may be affected by this action may file 

a request for a hearing and a petition to 

intervene with respect to issuance of the 

amendment to the subject facility 

operating license. 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 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland, 

and electronically on the Internet at the 

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

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

are problems in accessing the document, 

contact the PDR Reference staff at 1 

(800) 397–4209, (301) 415–4737, or by email to pdr.resource@nrc.gov. If a 

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to those specific 

sources and documents of which the 

petitioner is aware and on which the 

petitioner intends to rely to establish 

those facts or expert opinion. The 

petition must include sufficient 

information to show that a genuine 

dispute exists with the applicant on a 

material issue of law or fact.1

Contentions shall be limited to matters 

within the scope of the amendment 

under consideration. The contention 

must be one which, if proven, would 

entitle the petitioner to relief. A 

requestor/petitioner who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party. 

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Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 27837 

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical—primarily concerns/ 

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental—primarily 

concerns/issues relating to matters 

discussed or referenced in the 

environmental analysis for the 

applications. 

3. Miscellaneous—does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/ 

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. If a requestor/petitioner 

seeks to adopt the contention of another 

sponsoring requestor/petitioner, the 

requestor/petitioner who seeks to adopt 

the contention must either agree that the 

sponsoring requestor/petitioner shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring requestor/petitioner a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

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 EFiling process requires participants to 

submit and serve all adjudicatory 

documents over the Internet, or in some 

cases to mail copies on electronic 

storage media. 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/applycertificates.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/esubmittals.html. Participants may 

attempt to use other software not listed 

on the Web site, but should note that the 

NRC’s E-Filing system does not support 

unlisted software, and the NRC Meta 

System Help Desk will not be able to 

offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through EIE, users will be 

required to install a Web browser plugin 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/ 

e-submittals.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 

e-mail 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/e-submittals.

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 first-class mail 

as of the time of deposit in the mail, or 

by courier, express mail, or expedited 

delivery service upon depositing the 

document with the provider of the 

service. A presiding officer, having 

granted an exemption request from 

using E-Filing, may require a participant 

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27838 Federal Register / Vol. 75, No. 95 / Tuesday, May 18, 2010 / Notices 

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. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

March 29, 2010, as supplemented by 

letters dated March 29 and April 26, 

2010. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3.3.2, ‘‘Engineered 

Safety Feature Actuation System 

(ESFAS) Instrumentation,’’ Condition J 

under function 6.g in TS Table 3.3.2–1. 

Function 6.g provides an auxiliary 

feedwater (AFW) start signal that is 

provided to the motor-driven AFW 

pumps in the event of a trip of both 

turbine-driven main feedwater (MFW) 

pumps. The licensee determined that 

the design and normal operation of the 

MFW pumps could result in a condition 

that does not conform to TS Table 3.3.2– 

1, function 6.g. Entry into Limiting 

Condition for Operation (LCO) 3.0.3 will 

be required; therefore, the TS change 

was needed to address this condition. 

The change to Condition J allows 

placing the two channels in a tripped 

condition on one MFW pump when 

placing the pump into service or 

removing the pump from service prior 

to resetting the MFW pump. With the 

revision to Condition J, the licensee will 

not require an entry into LCO 3.0.3. 

Specifically, the changes revised 

Condition J for ESFAS instrumentation 

function 6.g to read, ‘‘One or more Main 

Feedwater Pumps trip channel(s) 

inoperable,’’ made corresponding 

changes to Required Action J.1, and 

placed a Note above Required Actions 

J.1 and J.2 for consistency with the 

revised Condition. 

Date of issuance: May 5, 2010. 

Effective date: As of its date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: 196. 

Facility Operating License No. NPF– 

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes (75 FR 

19431; April 14, 2010). 

The supplemental letters dated March 

29 and April 26, 2010, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed NSHC determination 

as published in the Federal Register. 

The notice provided an opportunity to 

submit comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. The 

notice also provided an opportunity to 

request a hearing by June 14, 2010, but 

indicated that if the Commission makes 

a final NSHC determination, any such 

hearing would take place after issuance 

of the amendment. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated May 5, 

2010. 

Attorney for licensee: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Dated at Rockville, Maryland, this 6th day 

of May 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–11564 Filed 5–17–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0180] 

Notice of Availability of Draft NUREG– 

1800, Revision 2; ‘‘Standard Review 

Plan for Review of License Renewal 

Applications for Nuclear Power Plants’’ 

and Draft NUREG–1801, Revision 2; 

‘‘Generic Aging Lessons Learned 

(GALL) Report’’ 

AGENCY: Nuclear Regulatory 

Commission (NRC). 

ACTION: Issuance of draft NUREG–1800, 

‘‘Standard Review Plan for Review of 

License Renewal Applications for 

Nuclear Power Plants ’’and draft 

NUREG–1801, ‘‘Generic Aging Lessons 

Learned (GALL) Report’’ for public 

comment; and announcement of public 

workshop. 

SUMMARY: The NRC staff is issuing drafts 

of the revised NUREG–1800, ‘‘Standard 

Review Plan for Review of License 

Renewal Applications for Nuclear 

Power Plants’’ (SRP–LR); and the revised 

NUREG–1801, ‘‘Generic Aging Lessons 

Learned (GALL) Report’’ for public 

comment. These revised documents 

describe methods acceptable to the NRC 

staff for implementing the license 

renewal rule, Title 10, Code of Federal 

Regulations Part 54 (10 CFR Part 54), as 

well as techniques used by the NRC staff 

in evaluating applications for license 

renewals. These draft documents 

supersede the preliminary draft 

documents that were publicly 

announced and placed on NRC’s Web 

site at http://www.nrc.gov/reactors/

operating/licensing/renewal/guidance/

updated-guidance.html on December 

23, 2009. 

The NRC is also announcing a public 

workshop to facilitate gathering public 

comments on the drafts of these revised 

documents. The NRC is especially 

interested in stakeholder comments that 

will improve the safety, effectiveness, 

and efficiency of the license renewal 

process. There are situations where the 

draft GALL Report, Revision 2 includes 

changes that have been previously 

issued for public comments as part of 

the staff’s license renewal Interim Staff 

Guidance (ISG) process. In particular, 

the Aging Management Program (AMP) 

XI.M40, ‘‘Monitoring of Neutron 

Absorbing Materials Other Than 

Boraflex’’ and related Aging 

Management Review (AMR) line items 

were processed by ISG LR–ISG–2009– 

01. Public comments were elicited on 

the proposed AMP XI.M40 by 74 FRN 

62829 dated December 1, 2009. Public 

comments were received, evaluated by 

the staff, and the proposed AMP 

XI.M40, and AMR line items, were 

revised as determined necessary by the 

staff. Because the staff has previously 

sought and received public comments 

on draft AMP XI.M40, the staff is not 

seeking further comments on this AMP 

as part of this Federal Register Notice 

(FRN). AMP XI.M40, and related AMR 

line items, are considered final by the 

staff. They have been included in the 

draft GALL Report, Revision 2 for 

completeness. 

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