51326 Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 

3Requesters should note that the filing 

requirements of the NRC’s E-Filing Rule (August 28, 

2007; 72 FR 49139) apply to appeals of NRC staff 

determinations (because they must be served on a 

presiding officer or the Commission, as applicable), 

but not to the initial SUNSI request submitted to the 

NRC staff under these procedures. 

(2) The requester may challenge the 

NRC staff’s adverse determination by 

filing a challenge within 5 days of 

receipt of that determination with: (a) 

The presiding officer designated in this 

proceeding; (b) if no presiding officer 

has been appointed, the Chief 

Administrative Judge, or if he or she is 

unavailable, another administrative 

judge, or an administrative law judge 

with jurisdiction pursuant to 10 CFR 

2.318(a); or (c) if another officer has 

been designated to rule on information 

access issues, with that officer. 

H. Review of Grants of Access. A 

party other than the requester may 

challenge an NRC staff determination 

granting access to SUNSI whose release 

would harm that party’s interest 

independent of the proceeding. Such a 

challenge must be filed with the Chief 

Administrative Judge within 5 days of 

the notification by the NRC staff of its 

grant of access. 

If challenges to the NRC staff 

determinations are filed, these 

procedures give way to the normal 

process for litigating disputes 

concerning access to information. The 

availability of interlocutory review by 

the Commission of orders ruling on 

such NRC staff determinations (whether 

granting or denying access) is governed 

by 10 CFR 2.311.3

I. The Commission expects that the 

NRC staff and presiding officers (and 

any other reviewing officers) will 

consider and resolve requests for access 

to SUNSI, and motions for protective 

orders, in a timely fashion in order to 

minimize any unnecessary delays in 

identifying those petitioners who have 

standing and who have propounded 

contentions meeting the specificity and 

basis requirements in 10 CFR Part 2. 

Attachment 1 to this Order summarizes 

the general target schedule for 

processing and resolving requests under 

these procedures. 

It is so ordered. 

Dated at Rockville, Maryland, this 30th day 

of September 2009. 

For the Nuclear Regulatory Commission. 

Annette L. Vietti-Cook, 

Secretary of the Commission. 

ATTACHMENT 1—General Target 

Schedule for Processing and Resolving 

Requests for Access to Sensitive 

Unclassified Non-Safeguards 

Information in This Proceeding 

Day Event/activity 

0 ......... Publication of Federal Register notice of hearing and opportunity to 

petition for leave to intervene, including order with instructions for 

access requests. 

10 ....... Deadline for submitting requests for 

access to Sensitive Unclassified 

Non-Safeguards Information 

(SUNSI) with information: Supporting the standing of a potential 

party identified by name and address; describing the need for the 

information in order for the potential party to participate meaningfully in an adjudicatory proceeding. 

60 ....... Deadline for submitting petition for 

intervention containing: (i) Demonstration of standing; (ii) all contentions whose formulation does 

not require access to SUNSI (+25 

Answers to petition for intervention; +7 petitioner/requestor 

reply). 

20 ....... Nuclear Regulatory Commission 

(NRC) staff informs the requester 

of the staff’s determination whether the request for access provides a reasonable basis to believe standing can be established 

and shows need for SUNSI. 

(NRC staff also informs any party 

to the proceeding whose interest 

independent of the proceeding 

would be harmed by the release 

of the information.) If NRC staff 

makes the finding of need for 

SUNSI and likelihood of standing, 

NRC staff begins document processing (preparation of redactions 

or review of redacted documents). 

25 ....... If NRC staff finds no ‘‘need’’ or no 

likelihood of standing, the deadline for petitioner/requester to file 

a motion seeking a ruling to reverse the NRC staff’s denial of 

access; NRC staff files copy of 

access determination with the 

presiding officer (or Chief Administrative Judge or other designated officer, as appropriate). If 

NRC staff finds ‘‘need’’ for 

SUNSI, the deadline for any party 

to the proceeding whose interest 

independent of the proceeding 

would be harmed by the release 

of the information to file a motion 

seeking a ruling to reverse the 

NRC staff’s grant of access. 

30 ....... Deadline for NRC staff reply to motions to reverse NRC staff determination(s). 

Day Event/activity 

40 ....... (Receipt +30) If NRC staff finds 

standing and need for SUNSI, 

deadline for NRC staff to complete information processing and 

file motion for Protective Order 

and draft Non-Disclosure Affidavit. Deadline for applicant/licensee to file Non-Disclosure 

Agreement for SUNSI. 

A ......... If access granted: Issuance of presiding officer or other designated 

officer decision on motion for protective order for access to sensitive information (including 

schedule for providing access 

and submission of contentions) or 

decision reversing a final adverse 

determination by the NRC staff. 

A + 3 .. Deadline for filing executed NonDisclosure Affidavits. Access provided to SUNSI consistent with 

decision issuing the protective 

order. 

A + 28 Deadline for submission of contentions whose development depends upon access to SUNSI. 

However, if more than 25 days 

remain between the petitioner’s 

receipt of (or access to) the information and the deadline for filing 

all other contentions (as established in the notice of hearing or 

opportunity for hearing), the petitioner may file its SUNSI contentions by that later deadline. 

A + 53 (Contention receipt +25) Answers to 

contentions whose development 

depends upon access to SUNSI. 

A + 60 (Answer receipt +7) Petitioner/Intervener reply to answers. 

>A + 60 Decision on contention admission. 

[FR Doc. E9–24049 Filed 10–5–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0433] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC) 

is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 


jlentini on DSKJ8SOYB1PROD with NOTICES
Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 51327 

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 September 

10, 2009, to September 23, 2009. The 

last biweekly notice was published on 

September 22, 2009 (74 FR 48316). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.92, 

this means that operation of the facility 

in accordance with the proposed 

amendment would not (1) involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rulemaking and 

Directives Branch (RDB), 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 RDB 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 petitioner/ 

requestor 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 petitioner/requestor 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/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor 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 petitioner/ 

requestor to relief. A petitioner/ 

requestor 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 


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51328 Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 

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, 

which the NRC promulgated in August 

28, 2007 (72 FR 49139). 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 

petitioner/requestor should contact the 

Office of the Secretary by e-mail at 

hearing.docket@nrc.gov, or by calling 

(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/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

representative) already holds an NRCissued digital ID certificate). Each 

petitioner/requestor will need to 

download the Workplace Forms 

ViewerTM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

is available at http://www.nrc.gov/sitehelp/e-submittals/install-viewer.html. 

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. 

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it 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 filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

the EIE 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 

EIE 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 through the 

‘‘Contact Us’’ link located on the NRC 

Web site at http://www.nrc.gov/sitehelp/e-submittals.html or by calling the 

NRC Meta-System Help Desk, which is 

available between 8 a.m. and 8 p.m., 

Eastern Time, Monday through Friday, 

excluding government holidays. The 

Meta-System Help Desk can be 

contacted by telephone at 1–866–672– 

7640 or by e-mail at 

MSHD.Resource@nrc.gov. 

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. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

the Atomic Safety and Licensing Board 

that the request and/or petition should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 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, an Atomic 

Safety and Licensing Board, or a 

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 

submissions. 

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 email to 

pdr.resource@nrc.gov. 

Dairyland Power Cooperative, Docket 

No. 50–409, La Crosse Boiling Water 

Reactor, Genoa, Wisconsin (TAC 

J00359) 

Date of amendment request: July 28, 

2009. 

Description of amendment request: 

The amendment application proposes 

changes to Technical Specifications, in 

support of the dry cask storage project 

at La Crosse Boiling Water Reactor. The 

application specifically proposes lower 

Fuel Element Storage Well water level 

limits and proposes changes to the 

definition of ‘‘fuel handling.’’ 

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? No. 

The proposed change to the definition of 

FUEL HANDLING is an administrative 


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Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 51329 

clarification and does not affect the operation 

of the plant or the postulated accidents in 

any way. The proposed changes to allow 

lower Fuel Element Storage Well (FESW) 

water level limits do not alter the manner in 

which individual fuel assemblies are moved 

or alter the design function of the FESW or 

any other structures, systems, and 

components used to ensure safe fuel storage. 

The total number of fuel assembly moves to 

the Dry Cask Storage System is exactly the 

same as that contemplated during original 

plant design when fuel was assumed to be 

transported from the plant directly to a 

disposal site. All of the accidents previously 

evaluated in the La Crosse Boiling Water 

Reactor (LACBWR) Decommissioning Plan 

have been reviewed for impact as a result of 

the proposed water level changes. The 

proposed changes do not affect the plant in 

such a manner that the likelihood or 

consequences of any previously evaluated 

accident is increased. 

Therefore, the proposed amendment 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? No. 

The proposed change to the definition of 

FUEL HANDLING is an administrative 

clarification and does not affect the operation 

of the plant in any way. The proposed 

changes to allow lower FESW water level 

limits do not alter the manner in which 

individual fuel assemblies are moved; or alter 

the design function of the FESW or any other 

structures, systems, and components used to 

ensure safe fuel storage. All of the accidents 

previously evaluated in the LACBWR 

Decommissioning Plan have been reviewed 

for impact as a result of the proposed water 

level changes. The existing accidents remain 

applicable and bounding for the LACBWR 

facility with the proposed changes in place 

and do not affect the plant in such a manner 

that a new accident has been created. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. 

The proposed change to the definition of 

FUEL HANDLING is an administrative 

clarification and does not affect plant 

operation or safety margins in any way. The 

proposed changes to allow lower FESW 

water level limits do not alter the manner in 

which individual fuel assemblies are moved; 

or alter the design function of the FESW or 

any other structures, systems, and 

components used to ensure safe fuel storage. 

All of the accidents previously evaluated in 

the LACBWR Decommissioning Plan have 

been reviewed for impact as a result of the 

proposed water level changes. The likelihood 

and consequences of previously evaluated 

accidents remain applicable and bounding 

with the proposed changes in place; thus, 

safety margins remain the same. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The U.S. Nuclear Regulatory 

Commission (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. 

NRC Branch Chief: Andrew Persinko. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit No. 3, 

Westchester County, New York 

Date of amendment request: July 23, 

2009. 

Description of amendment request: 

The proposed amendment would 

remove the level indicating instrument 

from the Technical Specification 

Surveillance Requirement (SR) for the 

refueling water storage tank, but leave 

the low level alarm function in the SR. 

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? 

No. The proposed change revises the 

existing Indian Point 3 Refueling Water 

Storage Tank (RWST) Technical 

Specification (TS) Surveillance Requirement 

(SR) to remove the level indication function 

for the L–921 instrument loop. Removal of a 

TS SR for the level indication does not 

increase the probability of an accident 

occurring since it is not an accident initiator 

and does not increase the consequences of an 

accident since it is not performing any 

mitigating function and is not a post accident 

instrument. The proposed revision will not 

affect RWST lo-lo level alarm function used 

for operator guidance to begin sequencing to 

Recirculation Mode of Safety Injection during 

a postulated loss of coolant accident (LOCA). 

There will be no change in equipment 

qualification requirements or changes to the 

surveillance requirement for the lo-lo level 

alarm. Therefore the proposed change does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed change removes the 

RWST level indication function from the 

RWST lo-lo level alarm surveillance 

requirement for the L–921 instrument loop. 

The proposed change does not involve 

installation of new equipment or 

modification of existing equipment, so that 

no new equipment failure modes are 

introduced. Also, the proposed change does 

not result in a change to the way that the 

equipment or facility is operated so that no 

new accident initiators are created. Therefore 

the proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The proposed change removes the 

RWST level indication function from the 

RWST io-lo level alarm surveillance 

requirement for the L–921 instrument loop. 

There is no change to the design 

requirements or the surveillance interval. 

The proposed change does not add the level 

indicating function elsewhere in the TS 

because it is a local level indication that is 

only used during normal operation and was 

never a post accident monitoring instrument. 

Therefore the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant (JAFNPP), Oswego 

County, New York 

Date of amendment request: July 31, 

2009. 

Description of amendment request: 

The proposed change would revise the 

JAFNPP Technical Specifications (TSs) 

Surveillance Requirements (SR) for 

testing of the Residual Heat Removal 

(RHR) System Shutdown Cooling (SDC) 

mode Containment Isolation, Reactor 

Pressure—High Function by replacing 

the current requirement to perform TS 

SR 3.3.6.1.3, Perform Channel 

Calibration, with TS SR 3.3.6.1.1 

Perform Channel Check, SR 3.3.6.1.2, 

Perform Channel Functional Test, SR 

3.3.6.1.4, Calibrate the Trip Units, and 

SR 3.3.6.1.5, Perform Channel 

Calibration. These changes are to 

support a proposed plant modification 

to increase the reliability of SDC 

isolation logic by changing the source of 

the reactor high pressure input signal. 

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: 


jlentini on DSKJ8SOYB1PROD with NOTICES
51330 Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 

1. Will operation of the facility in 

accordance with this proposed change 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

Response: No. 

The proposed change modifies the SRs that 

demonstrate the operability of the SDC 

Isolation, Reactor Pressure—High function. 

The current surveillance requirements 

include a 92-day calibration and a 24-month 

logic system functional test. These 

surveillance requirements are typical for 

pressure switches installed on dedicated 

process measurement lines. The proposed 

change in surveillance requirements is 

consistent with the use of ATTS [Analog 

Transmitter Trip System] transmitters 

installed on shared process measurement 

lines. The proposed surveillance 

requirements include the standard 

requirements applied to all ATTS equipment 

and thus will result in acceptable 

demonstration of the operability of the SDC 

Isolation Reactor Pressure—High function. 

The ATTS equipment that will be used for 

the SDC Isolation, Reactor Pressure—High 

function is classified as safety related and is 

environmentally qualified. The logic input 

configuration of the ATTS equipment will be 

the same as the configuration of the pressure 

switches. This will assure the same 

functionality currently performed by the 

pressure switches currently used for the SDC 

Isolation Reactor Pressure—High function. 

The reliability of the ATTS has been proven 

in other RPS [Reactor Protection System], 

PCIS [Primary Containment Isolation 

System], and ECCS [Emergency Core Cooling 

System] functions and is comparable to the 

reliability of the pressure switches that 

currently perform the SDC Isolation, Reactor 

Pressure—High function. Therefore, the 

consequences of any accident mitigated by 

the SDC Isolation, Reactor Pressure—High 

function will not increase. 

Based on these considerations, the 

proposed surveillance requirement changes 

do not involve a significant increase in the 

probability or consequences of an accident 

’previously evaluated. 

2. Will operation of the facility in 

accordance with this proposed change create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change aligns the TS 

surveillance requirements with the type of 

equipment that will be used to supply the 

reactor pressure input to the SDC Isolation 

Reactor Pressure—High logic. Since the 

transmitters that will be used to supply the 

reactor pressure input are currently installed 

equipment there are no new accidents 

introduced by the equipment. The proposed 

change in SRs aligns the requirements with 

the—requirements currently imposed on the 

equipment in other JAF TS applications. The 

performance of the SDC Isolation, Reactor 

Pressure—High function, is not altered by 

changing the input source for reactor 

pressure parameter. Redundant power 

sources within the ATTS assure the 

functionality of the system during all plant 

operating modes that require the SDC 

Isolation, Reactor Pressure—High function. 

The proposed change will not introduce any 

new failure modes and, therefore, does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Will operation of the facility in 

accordance with this proposed change 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The TS surveillance requirements that will 

be imposed on the SDC Isolation, Reactor 

Pressure—High function reflect the 

equipment that will perform that function. 

The proposed change in surveillance 

requirements will appropriately demonstrate 

the operability of the SDC Isolation, Reactor 

Pressure—High function. 

Since the proposed changes to the SRs are 

consistent with the SRs for ATTS 

transmitters in other RPS, PCIS, and ECCS 

applications the proposed requirements have 

been demonstrated to provide an adequate 

margin of safety. Therefore, the proposed 

change does not involve a significant 

reduction in any 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: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Operations, Inc., Docket No. 50– 

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request: August 5, 

2009. 

Description of amendment request: 

Current Technical Specification (TS) 

5.5.8, ‘‘Inservice Testing Program,’’ 

contains references to the American 

Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel 

Code, Section XI as the source of 

requirements for the inservice testing 

(IST) of ASME Code Class 1, 2, and 3 

pumps and valves. The proposed 

amendment would delete the references 

to Section XI of the Code and 

incorporate references to the ASME 

Code for Operation and Maintenance of 

Nuclear Power Plants (ASME OM Code). 

The proposed amendment would also 

indicate that there may be some 

nonstandard frequencies utilized in the 

IST Program in which the provisions of 

Surveillance Requirement 3.0.2 are 

applicable. The proposed changes are 

consistent with Technical Specification 

Task Force (TSTF) Technical Change 

Travelers 479–A, ‘‘Changes to Reflect 

Revision to 10 CFR 50.55a,’’ and 497– 

A, ‘‘Limit Inservice Testing Program SR 

3.0.2 Application to Frequencies of 2 

Years or Less.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises TS 5.5.8, 

Inservice Testing Program, for consistency 

with the requirements of 10 CFR 50.55a(f)(4) 

for pumps and valves which are classified as 

American Society of Mechanical Engineers 

(ASME) Code Class 1, Class 2 and Class 3. 

The proposed change incorporates revisions 

to the ASME Code which is consistent with 

the expectations of 10 CFR 50.55(a). 

The proposed change does not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. The proposed change does not 

involve the addition or removal of any 

equipment, or any design changes to the 

facility. Therefore, this proposed change does 

not represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

modification to the physical configuration of 

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

installed) or change in the methods 

governing normal plant operation. The 

proposed change does not introduce a new 

accident initiator, accident precursor, or 

malfunction mechanism. Therefore, this 

proposed change does not create the 

possibility of an accident or a different kind 

than previously evaluated. 

3. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises TS 5.5.8, 

Inservice Testing Program, for consistency 

with the requirements of 10 CFR 50.55a(f)(4) 

for pumps and valves which are classified as 

ASME Code Class 1, Class 2 and Class 3. The 

proposed change incorporates revisions to 

the ASME Code, which is consistent with the 

expectations of 10 CFR 50.55a. The safety 

function of the affected pumps and valves are 

maintained. Therefore, this proposed change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 


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amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Terence A. 

Burke, Associate General Council— 

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Michael T. 

Markley. 

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 amendment request: July 30, 

2009. 

Description of amendment request: 

The proposed amendment would delete 

Technical Specification (TS) Section 

3.6.3.1, ‘‘Containment Atmosphere 

Dilution (CAD) System,’’ to modify 

containment combustible gas control 

requirements as permitted by Title 10 of 

the Code of Federal Regulations, Part 50 

Section 50.44 (10 CFR 50.44). 10 CFR 

50.44 was revised on September 16, 

2003, as noticed in the Federal Register 

(68 FR 54123). 

The Nuclear Regulatory Commission 

(NRC) staff issued a ‘‘Notice Of 

Opportunity To Comment On Model 

Safety Evaluation, Model No Significant 

Hazards Determination, And Model 

Application For Licensees that Wish To 

Adopt TSTF–478, Revision 2, ‘BWR 

[Boiling-Water Reactor] Technical 

Specification Changes that Implement 

the Revised Rule for Combustible Gas 

Control,’’ in the Federal Register on 

October 11, 2007 (72 FR 57970). The 

notice included a model safety 

evaluation (SE) and a model no 

significant hazards consideration 

(NSHC) determination. On November 

21, 2007, the NRC staff issued a notice 

in the Federal Register (72 FR 65610) 

announcing that the model SE and 

model NSHC determination may be 

referenced in plant-specific applications 

to adopt the changes. In its application 

dated July 30, 2009, the licensee 

affirmed the applicability of the model 

NSHC determination which is presented 

below. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC 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 Containment Atmosphere Dilution 

(CAD) system is not an initiator to any 

accident previously evaluated. The TS 

Required Actions taken when a drywell 

cooling system fan is inoperable are not 

initiators to any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. 

The revised 10 CFR 50.44 no longer defines 

a design-basis accident (DBA) hydrogen 

release and the Commission has 

subsequently found that the DBA loss-ofcoolant accident (LOCA) hydrogen release is 

not risk significant. In addition, CAD has 

been determined to be ineffective at 

mitigating hydrogen releases from the more 

risk significant beyond DBAs that could 

threaten containment integrity. Therefore, 

elimination of the CAD system will not 

significantly increase the consequences of 

any accident previously evaluated. The 

consequences of an accident while relying on 

the revised TS Required Actions for drywell 

cooling system fans are no different than the 

consequences of the same accidents under 

the current Required Actions. As a result, the 

consequences of any accident previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Criterion 2: The proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated 

No new or different accidents result from 

utilizing the proposed change. The proposed 

change permits physical alteration of the 

plant involving removal of the CAD system. 

The CAD system is not an accident precursor, 

nor does its existence or elimination have 

any adverse impact on the pre-accident state 

of the reactor core or post accident 

confinement of radionuclides within the 

containment building from any design basis 

event. The changes to the TS do not alter 

assumptions made in the safety analysis, but 

reflect changes to the design requirements 

allowed under the revised 10 CFR 50.44. The 

proposed change is consistent with the 

revised safety analysis assumptions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Criterion 3: The proposed change does not 

involve a significant reduction in a margin of 

safety 

The Commission has determined that the 

DBA LOCA hydrogen release is not risk 

significant, therefore is not required to be 

analyzed in a facility accident analysis. The 

proposed change reflects this new position 

and, due to remaining plant equipment, 

instrumentation, procedures, and programs 

that provide effective mitigation of and 

recovery from reactor accidents, including 

postulated beyond design basis events, does 

not result in a significant reduction in a 

margin of safety. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

Based on the above, the NRC 

concludes that the proposed change 

presents no significant hazards 

consideration under the standards set 

forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

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: Mr. J. Bradley 

Fewell, Associate General Counsel, 

Exelon Generation Company LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station (FCS), 

Unit No. 1, Washington County, 

Nebraska 

Date of amendment request: May 29, 

2009. 

Description of amendment request: 

The proposed amendment would: (1) 

Revise the definition for OperableOperability in the FCS Technical 

Specifications (TS); (2) modify the 

provisions under which equipment may 

be considered operable when either its 

normal or emergency power source is 

inoperable; (3) delete TS limiting 

condition for operation (LCO) 2.0.1(2); 

(4) delete diesel generator surveillance 

requirement (SR) 3.7(1)e; and (5) 

relocate the guidance for inoperable 

power supplies and verifying 

operability of redundant components 

into the LCO for electrical equipment 

2.7, Electrical Systems. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to revise the 

definition of operable-operability, modify the 

provisions under which equipment may be 

considered operable when either its normal 

or emergency power source is inoperable, 

delete Technical Specification (TS) limiting 

conditions for operation (LCO) 2.0.1(2), and 

relocate the guidance for inoperable power 

supplies and verifying operability of 

redundant components into the LCO for 

electrical equipment is more aligned with 

NUREG–1432, Standard Technical 

Specifications [STS] for Combustion 

Engineering Plants, and does not adversely 

impact the probability of an accident 


jlentini on DSKJ8SOYB1PROD with NOTICES
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previously evaluated. The proposed changes 

are being made to address inconsistencies in 

guidance provided in TS 2.0.1(2) and TS 

2.7(2). The proposed change does not affect 

the operability requirements for the 

emergency diesel generators (EDGs) or the 

house service transformers, and therefore 

does not impact the consequences of an 

analyzed accident. 

The new requirement added to TS 2.7 

provides assurance that a loss of offsite 

power during the period that an EDG (or 

house service transformer) is inoperable, or 

loss of an EDG during the period that a house 

service transformer is inoperable, or loss of 

a house service transformer during the period 

that an EDG is inoperable, does not result in 

a complete loss of safety function of critical 

systems; thereby such a loss does not 

significantly increase the probability of an 

accident. 

Consistent with NUREG 1432, the 4-hour 

allowed time added to TS 2.7(2)j for the 

EDGs, takes into account the capacity and 

capability of the remaining alternating 

current (AC) sources, a reasonable time for 

repairs, and the low probability of a design 

basis accident (DBA) occurring during this 

period. On a component basis, single failure 

protection for the required feature’s function 

may have been lost; however, function has 

not been lost. 

Additionally, consistent with NUREG– 

1432, the 24-hour allowed time added to TS 

2.7(2)b for the house service transformers 

takes into account the capacity and capability 

of the remaining AC sources, a reasonable 

time for repairs, and the low probability of 

a DBA occurring during this period. 

The proposed change removes the 

surveillance requirement (SR) to perform an 

inspection of the EDG on a refueling 

inspection frequency in accordance with the 

manufacturer’s recommendations. This 

inspection is considered a maintenance 

activity, not an SR, and has no impact on the 

probability of an accident since EDGs are not 

initiators for any analyzed event. Deletion of 

TS SR 3.7(1)e from the TS does not impact 

the capability of the EDGs to perform their 

accident mitigation functions. The required 

EDG maintenance inspections will continue 

to be performed in accordance with the 

licensee-controlled EDG maintenance 

process. The consequences of an accident are 

not impacted because EDG operability is 

controlled by other portions of TS 3.7, which 

ensures that required surveillances are 

performed. The appropriate LCOs are entered 

in the event that EDG surveillance criteria are 

not met. 

As a result of redefining ‘‘OPERABLE’’ and 

adding the provision to TS 2.7(2)j, the 

statements ‘‘provided there are no inoperable 

required engineered safeguards components 

which are redundant’’ related to the electrical 

distribution components are being deleted 

from the other 2.7(2) TS for the buses, 

transformer, and motor control center (MCC) 

for clarification and consistency because 

these statements restrict only to engineered 

safeguards components. In addition, the 

administrative changes to renumber the 

existing TS sections ‘‘TS 2.0.1(3) to 2.0.1(2)’’ 

and TS 3.7(1)f to TS 3.7(1)e. are being made 

as a result of deletions to previous TS 

paragraphs and are being made for 

consistency and clarification. Rearranging the 

listing order of the MCCs in TS 2.7(1)f and 

TS 2.7(2)g in bus order clarifies the TS. As 

such, these editorial changes are not 

initiators of any accidents previously 

evaluated. As a result, the probability of an 

accident previously evaluated is not affected. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

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

or different type of equipment will be 

installed) or a change in methods governing 

normal plant operation. The proposed 

changes to TS 2.0.1(2) and TS 2.7 do not 

create the possibility of a new or different 

kind of accident since the design function of 

the affected equipment is not changed. No 

new interactions between systems or 

components are created. No new failure 

mechanisms of associated systems will exist. 

By deleting TS LCO 2.0.1(2) and including 

the guidance in TS 2.7, inconsistencies in the 

existing TS will be eliminated. The new 

requirements added to TS 2.7 will include 

guidance to declare required systems or 

components without a normal or emergency 

power source available inoperable, when a 

redundant system or component is also 

inoperable. This provides assurance that a 

loss of offsite power, during the period that 

an EDG (or house service transformer) is 

inoperable, or loss of an EDG during the 

period that a house service transformer is 

inoperable (or vice versa), does not result in 

a complete loss of safety function of critical 

systems. 

No new failure mechanisms would be 

created. The proposed changes do not alter 

any assumptions made in the safety analyses. 

For the most part, the proposed changes are 

more aligned with the STS. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes to delete TS 2.0.1(2) 

and relocate the guidance for inoperable 

power supplies and verifying operability of 

redundant components to TS LCO 2.7(2)j, to 

delete the statement that MCC–3C1 may be 

inoperable in excess of 8 hours if battery 

chargers No. 1 and No. 2 are operable, and 

to delete the SR for inspecting the DG on a 

refueling frequency in accordance with the 

manufacturer’s recommendations do not alter 

the manner in which safety limits or limiting 

safety system settings are determined. The 

safety analysis acceptance criteria are not 

affected by these proposed changes. The 

sources of power credited for design basis 

events are not affected by the proposed 

changes. 

The proposed changes to modify the 

provisions under which equipment may be 

considered operable when either its normal 

or emergency power source is inoperable, 

delete TS LCO 2.0.1(2), and relocate the 

guidance for inoperable power supplies and 

verifying operability of redundant 

components into the LCO for electrical 

equipment is more aligned with the STS. 

These changes are being made to address 

inconsistencies in guidance provided in TS 

2.0.1(2) and TS 2.7(2). The proposed change 

does not reduce the operability requirements 

for the transformers, buses, MCCs, or EDGs 

and therefore will not result in plant 

operation in a configuration outside of the 

design basis. 

Further, the proposed change does not 

change the design function of any equipment 

assumed to operate in the event of an 

accident. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David A. Repka, 

Esq., Winston & Strawn, 1700 K Street, 

NW., Washington, DC 20006–3817. 

NRC Branch Chief: Michael T. 

Markley. 

PPL Susquehanna, LLC, Docket Nos. 50– 

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 

Units 1 and 2), Luzerne County, 

Pennsylvania 

Date of amendment request: March 

24, 2009, as supplemented by letters 

dated April 24, and September 11, 2009. 

Description of amendment request: 

The proposed change revises the 

allowable value in the Technical 

Specification (TS) Table 3.3.5.1–1 

(Function 3.d) for the high-pressure 

coolant injection (HPCI) automatic 

pump suction transfer from the 

condensate storage tank (CST) to the 

suppression pool (SP). The present 

allowable value for this transfer is 

greater than or equal to 36 inches above 

the CST bottom. The proposed change is 

to increase the allowable value for this 

transfer to occur at greater than or equal 

to 40.5 inches above the CST bottom. 

Additionally, the proposed 

amendment also includes an editorial/ 

administrative change which corrects a 

typographical error in the SSES Units 1 

and 2 TS Section 3.10.8.f. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 


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1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change to TS Table 

3.3.5.1–1 increases the Technical 

Specification allowable value for the HPCI 

suction low level automatic transfer function 

from ≥ 3 6 inches to ≥ 40.5 inches above the 

CST bottom. There are no process setpoint 

changes associated with this TS allowable 

value change. This TS change does not 

introduce the possibility of an increase in the 

probability or consequences of an accident 

because the HPCI automatic transfer function 

is not an initiator of any new accidents nor 

does it introduce any new failure modes. The 

CST is not safety related and therefore not 

credited in any design basis accident 

analyses. However, the CST reserve volume 

is credited in anticipated transients without 

scram (ATWS), Appendix R and station 

blackout (SBO) evaluations. The reserve 

volume available in the CST at the proposed 

allowable value of 40.5 inches above the CST 

bottom remains adequate to fully support 

these HPCI system support functions and the 

change fully supports HPCI system operation. 

The reserve volume is not reduced as a result 

of the proposed change in the TS allowable 

value since the transfer will still occur at the 

CST low level instrument setpoint of 43.5 

inches above tank bottom, which remains 

unchanged. 

The HPCI system automatic transfer 

function occurs at the point in a design basis 

accident (DBA) when the CST level reaches 

the low level transfer setpoint. This proposed 

change will require the HPCI pump suction 

to be transferred from the CST to the SP at 

40.5 inches versus 36 inches above the CST 

bottom. Currently, the TS allow this transfer 

to occur at 36 inches. This proposed change 

is conservative because it assures the suction 

transfer will occur while there is more water 

in the tank, thus eliminating the possibility 

of vortex formation and air intrusion to the 

HPCI pump suction. Since this proposed 

change ensures the HPCI system automatic 

suction transfer function occurs without 

adversely impacting HPCI system operation, 

it does not involve a significant increase in 

the probability or consequences of an 

accident previously evaluated. 

The proposed editorial/administrative 

change is necessary to correct a typographical 

error in the SSES Units 1 and 2 TS Section 

3.10.8.f. This editorial 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? 

No. As discussed above, the proposed 

change to TS Table 3.3.5.1–1 involves 

increasing the TS allowable value for the 

HPCI low level automatic transfer function 

from the CST to the SP at ≥ 36 inches to ≥ 

40.5 inches above the CST tank bottom. This 

change ensures the HPCI automatic transfer 

function occurs without introducing the 

possibility of vortex formation or air 

intrusion in the HPCI pump suction path. All 

HPCI system support functions remain 

unaffected by this change. This TS change 

does not introduce the possibility of a new 

accident because the HPCI automatic transfer 

function is not an initiator of any accident 

and no new failure modes are introduced. 

There are no new types of failures or new or 

different kinds of accidents or transients that 

could be created by these changes. Therefore, 

this change does not create the possibility of 

a new or different kind of accident from any 

accident previously evaluated. 

The proposed editorial/administrative 

change only corrects a typographical error in 

the SSES Units 1 and 2 TS Section 3.10.8.f. 

This editorial change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The margin of safety is established 

through equipment design, operating 

parameters, and the setpoints at which 

automatic actions are initiated. The proposed 

change to TS Table 3.3.5.1–1 involves 

increasing the allowable level at which the 

HPCI automatic suction transfer from the 

CST to the SP must occur to avoid the 

possibility of vortex formation or air 

intrusion into the HPCI pump. This change 

does not result in a change to the level switch 

setpoint, which initiates the HPCI suction 

transfer from the CST to the SP. Although the 

allowable value for the transfer is now closer 

to the process setpoint for activation of the 

level switch, this reduction in operating 

margin was reviewed and determined to be 

acceptable. The level switch setpoint 

tolerances were established based on 

historical instrument data and instrument 

characteristics. These tolerances provide 

adequate margin to the proposed TS 

allowable value of 40.5 inches above the CST 

bottom. The tolerances further ensure the 

transfer will occur prior to level reaching the 

technical specification allowable value. 

Therefore, the proposed change does not 

result in a significant reduction in a margin 

of safety. 

The proposed editorial/administrative 

change only corrects a typographical error in 

the SSES Units 1 and 2 TS Section 3.10.8.f. 

This editorial change does not result in 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: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Nancy L. Salgado. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: July 30, 

2009. 

Description of amendment request: 

The proposed amendment would 

relocate Technical Specification (TS) 

surveillance requirements (SRs) for the 

reactor recirculation system motorgenerator (MG) set scoop tube stop 

settings to the Technical Requirements 

Manual (TRM). Specifically, the 

proposed amendment would relocate TS 

SR 4.4.1.1.3 to the TRM which is a 

licensee-controlled document. SR 

4.4.1.1.3 requires that each MG set 

scoop tube mechanical and electrical 

stop be demonstrated operable with 

overspeed setpoints less than or equal to 

109% and 107%, respectively, of rated 

core flow, at least once per 18 months. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration. The Nuclear Regulatory 

Commission (NRC) staff’s review 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 major components in the MG set 

consist of a motor, fluid coupler and a 

generator. The motor drives the generator 

through the fluid coupler. The speed and 

output of the generator rise and fall as the 

volume of fluid in the coupler is varied by 

changing the position of the scoop tube. As 

the generator’s output increases or decreases, 

the speed of the recirculation pump follows 

suit. The scoop tube mechanism has both 

mechanical and electrical overspeed stops 

that limit recirculation flow by limiting the 

MG set speed. The electrical stop actuates 

first. The mechanical stop is designed to 

prevent the scoop tube motion if the 

electrical stop fails or to mitigate overshoot 

of the electrical stop. The electrical stops are 

not credited in any of the accident or 

transient analyses. The mechanical stop 

settings are an input used in the 

determination of the flow dependent 

minimum critical power ratio (MCPR) and 

the linear heat generation rate (LHGR) or 

average planar linear heat generation rate 

(APLHGR) operating limits. These operating 

limits are established and documented on a 

cycle-specific basis in the core operating 

limits report (COLR) in accordance with TS 

6.9.1.9. Operation within the MCPR, LGHR 

and APLHGR operating limits is required in 

accordance with TSs 3.2.3, 3.2.4, and 3.2.1, 

respectively. 

Once relocated, any future changes to the 

surveillance requirements for the MG set 

scoop tube mechanical and electrical stop 

settings would be controlled by 10 CFR 

50.59. 

There are no physical plant modifications 

associated with this change. The proposed 

amendment would not alter the way any 

structure, system, or component (SSC) 

functions and would not alter the way the 

plant is operated. As such, the proposed 


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amendment would have no impact on the 

ability of the affected SSCs to either preclude 

or mitigate an accident. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment would not 

change the design function or operation of 

the SSCs involved and would not impact the 

way the plant is operated. As such, the 

proposed change would not introduce any 

new failure mechanisms, malfunctions, or 

accident initiators not already considered in 

the design and licensing bases. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The margin of safety is associated with the 

confidence in the ability of the fission 

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

coolant pressure boundary, and containment 

structure) to limit the level of radiation to the 

public. There are no physical plant 

modifications associated with the proposed 

amendment. The proposed amendment 

would not alter the way any SSC functions 

and would not alter the way the plant is 

operated. The proposed amendment would 

not introduce any new uncertainties or 

change any existing uncertainties associated 

with any safety limit. The proposed 

amendment would have no impact on the 

structural integrity of the fuel cladding, 

reactor coolant pressure boundary, or 

containment structure. Based on the above 

considerations, the NRC staff concludes that 

the proposed amendment would not degrade 

the confidence in the ability of the fission 

product barriers to limit the level of radiation 

to the public. Therefore, the proposed change 

does not involve a significant reduction in a 

margin of safety. 

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: Vincent 

Zabielski, PSEG Nuclear LLC–N21, P.O. 

Box 236, Hancocks Bridge, NJ 08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Calvert Cliffs Nuclear Power Plant, Inc., 

Docket Nos. 50–317 and 50–318, Calvert 

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendments: 

May 21, 2009. 

Brief description of amendments: The 

amendments removed the Table of 

Contents from the Technical 

Specifications and place them under 

licensee control. 

Date of issuance: September 21, 2009. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment Nos.: 293 and 269. 

Renewed Facility Operating License 

Nos. DPR–53 and DPR–69: Amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 30, 2009 (74 FR 31320). 

The Commission’s related evaluation 

of these amendments is contained in a 

Safety Evaluation dated September 21, 

2009. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina. 

Duke Power Company LLC, Docket 

Nos. 50–369 and 50–370, McGuire 

Nuclear Station, Units 1 and 2, 

Mecklenburg County, North Carolina. 

Duke Power Company LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina. 

Date of application for amendments: 

February 27, 2009. 

Brief description of amendments: The 

amendments deleted those portions of 

the Technical Specifications (TSs) 

superseded by the Code of Federal 

Regulations, Part 26, Subpart I. The 

changes are consistent with Nuclear 

Regulatory Commission (NRC)-approved 

Revision 0 to Technical Specification 

Task Force (TSTF) Improved Standard 

Technical Specification Change 

Traveler, TSTF–511, ‘‘Eliminate 

Working Hour Restrictions from TS 

5.2.2 to Support Compliance with 10 

CFR Part 26.’’ 

Date of issuance: September 21, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 251 and 246. 

Renewed Facility Operating License 

Nos. NPF–35 and NPF–52: Amendments 

revised the licenses and technical 

specifications. 

Amendment Nos.: 253 and 233. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: Amendments 

revised the licenses and technical 

specifications. 

Amendment Nos.: 365, 367, and 366. 

Renewed Facility Operating License 

Nos. DPR–38, DPR–47, and DPR–55: 

Amendments revised the licenses and 

technical specifications. 

Date of initial notices in Federal 

Register: August 11, 2009 (74 FR 

40236) Catawba and McGuire; and 

August 11, 2009 (74 FR 40237) Oconee. 

The Commission’s related evaluation 

and final finding of no significant 

hazards consideration of the 


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amendments is contained in a Safety 

Evaluation dated September 21, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment: 

November 13, 2008, as supplemented by 

letters dated June 1, July 14, and August 

17, 2009. 

Brief description of amendment: The 

amendment modified Technical 

Specification 3.3.1.1, Reactor Protective 

Instrumentation, specifically Table 4.3– 

1 and associated Notes 7 and 8, to 

clarify and streamline Reactor Coolant 

System flow verification requirements 

associated with the Departure from 

Nucleate Boiling Ratio reactor trip 

signal. 

Date of issuance: September 16, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 286. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in Federal 

Register: January 27, 2009 (74 FR 

4769). The supplemental letters dated 

June 1, July 14, and August 17, 2009, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 16, 

2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment: 

May 13, 2009, as supplemented by letter 

dated July 8, 2009. 

Brief description of amendment: The 

amendment modified Technical 

Specification 2.1.1.1, departure from 

nucleate boiling ratio safety limit based 

upon the Combustion Engineering 16 $× 

16 Next Generation Fuel design and the 

associated departure from nucleate 

boiling correlations. 

Date of issuance: September 18, 2009. 

Effective date: As of the date of 

issuance and shall be implemented after 

the current cycle (Cycle 20) is 

completed and prior to startup for 

operating Cycle 21. 

Amendment No.: 287. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in Federal 

Register: June 30, 2009 (74 FR 31321). 

The supplemental letter dated July 8, 

2009, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 18, 

2009. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of application for amendments: 

July 25, 2008 (Agencywide Documents 

Access and Management System 

(ADAMS) Accession No. 

ML082110187), as supplemented by 

letters dated October 31, 2008 (ADAMS 

Accession No. ML083080059), February 

17, 2009 (ADAMS Accession No. 

ML090480372), May 8, 2009 (ADAMS 

Accession No. ML092380433) and July 

27, 2009 (ADAMS Accession No. 

ML092100162). 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) 3.3.1.1, ‘‘Reactor 

Protection System (RPS) 

Instrumentation,’’ Surveillance 

Requirement (SR) 3.3.1.1.8 and TS 

3.3.1.3, ‘‘Oscillation Power Range 

Monitor (OPRM) Instrumentation,’’ SR 

3.3.1.3.2 to increase the frequency 

interval between Local Power Range 

Monitor calibrations from 1000 effective 

full power hours (EFPH) to 2000 EFPH. 

Date of issuance: September 16, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 195/182. 

Facility Operating License Nos. NPF– 

11 and NPF–18: The amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: January 23, 2009 (74 FR 

4250–4251). The October 31, 2008, 

February 17, 2009, May 8, 2009, and 

July 27, 2009 supplements, contained 

clarifying information and did not 

change the NRC staff’s initial proposed 

finding of no significant hazards 

consideration nor expand the scope of 

the original Federal Register notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 16, 

2009. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1 (TMI–1), 

Dauphin County, Pennsylvania 

Date of application for amendment: 

October 9, 2008, supplemented by letter 

dated April 2, 2009. 

Brief description of amendment: The 

amendment reflects the planned 

installation of replacement steam 

generators (SGs). Specifically, the 

amendment modified the technical 

specifications to eliminate the existing 

requirements associated with tube 

sleeve repairs and alternate repair 

criteria which are not applicable to the 

replacement SGs. It also incorporated a 

revised primary-to-secondary leakage 

criterion, changes the required reporting 

period for SG inspection results, and 

incorporated revised tube integrity 

surveillance frequency requirements to 

reflect the new Alloy 690 tubing 

material. 

Date of issuance: September 15, 2009. 

Effective date: Upon installation of 

the replacement SGs and shall be 

implemented prior to exiting cold 

shutdown from the TMI–1 SG 

replacement refueling outage (T1R18), 

which is scheduled to begin in the fall 

of 2009. 

Amendment No.: 271. 

Facility Operating License No. DPR– 

50: Amendment revised the license and 

the technical specifications. 

Date of initial notice in Federal 

Register: March 10, 2009 (74 FR 

10310). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 15, 

2009. 

No significant hazards consideration 

comments received: No. 

FPL Energy Duane Arnold, LLC, Docket 

No. 50–331, Duane Arnold Energy 

Center, Linn County, Iowa 

Date of application for amendment: 

January 30, 2009, as supplemented by a 

letter dated July 30, 2009. 

Brief description of amendment: The 

amendment deleted the Duane Arnold 

Energy Center (DAEC) Technical 

Specification (TS) Section 5.2.2.e 

regarding work hour controls. 

Date of issuance: September 18, 2009. 

Effective date: As of the date of 

issuance and shall be implemented by 

October 1, 2009. Amendment No.: 274. 


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Facility Operating License No. DPR– 

49: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 24, 2009 (74 FR 

12393). The supplemental letter 

contained clarifying information, did 

not change the initial no significant 

hazards consideration determination, 

and did not expand the scope of the 

original Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 18, 

2009. 

No significant hazards consideration 

comments received: No. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: October 

13, 2008, as supplemented by letters 

dated April 8, May 29, June 12, and 

September 1, 2009. 

Brief description of amendment: The 

amendment revised the licensing basis 

by approving adoption of the 

Alternative Source Term (AST), in 

accordance with Section 50.67 of Title 

10 of the Code of Federal Regulations 

(10 CFR), for use in calculating the lossof-coolant accident (LOCA) dose 

consequences. The amendment revised 

the Technical Specifications (TSs) to (1) 

change the TS definition for DOSE 

EQUIVALENT I–131 to adopt Federal 

Guidance Report 11 dose conversion 

factors; (2) require operability of the 

Standby Liquid Control system in Mode 

3, to reflect its credit in the LOCA 

analysis; (3) establish a Main Steam 

(MS) Pathway leakage limit that 

effectively increases the previous MS 

isolation valve leakage limit; and (4) 

change TS Section 5.5.12 to reflect a 

requested permanent exemption from 

the requirements of 10 CFR Part 50, 

Appendix J, Option B, Section III.A, to 

allow exclusion of MS Pathway leakage 

from the overall integrated leakage rate 

measured during the performance of a 

Type A test, and from the requirements 

of Appendix J, Option B, Section III.B, 

to allow exclusion of the MS Pathway 

leakage from the combined leakage rate 

of the penetrations and valves subject to 

Type B and C tests. 

Date of issuance: September 15, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days of issuance. 

Amendment No.: 234. 

Facility Operating License No. DPR– 

46: Amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 23, 2009 (74 FR 

4251). The supplemental letters dated 

April 8, May 29, June 12, and September 

1, 2009, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 15, 

2009. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: March 

30, 2009. 

Brief description of amendment 

request: The amendments revised 

Technical Specification (TS) by deleting 

the Reactor Coolant Pump breaker 

position reactor trip in TS 3.3.1, 

‘‘Reactor Trip System (RTS) 

Instrumentation.’’ 

Date of Issuance: September 18, 2009. 

Amendment Nos.: Unit 1–183; Unit 

2–176. 

Facility Operating License Nos. NPF– 

2 and NPF–8: The amendment revised 

the Facility Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: May 19, 2009 (74 FR 23448). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 18, 

2009. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–390, Watts Bar Nuclear Plant (WBN), 

Unit 1, Rhea County, Tennessee 

Date of application for amendment: 

June 5, 2009. 

Brief description of amendment: The 

amendment revised WBN Unit 1 

technical specifications (TSs) to revise 

the completion time from 1 hour to 24 

hours for Condition B of TS 3.5.1, 

‘‘Accumulators’’ and its associated 

Bases. 

Date of issuance: September 9, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days of issuance. 

Amendment No.: 81. 

Facility Operating License No. NPF– 

90: Amendment revises TS 3.5.1. 

Date of initial notice in Federal 

Register: June 30, 2009 (74 FR 31326). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 9, 

2009. 

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 

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 


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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. 

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 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/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 email 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 O1F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland, and electronically on the 

Internet at the NRC Web site, http:// 

www.nrc.gov/reading-rm/doccollections/cfr/. If there are problems in 

accessing the document, contact the 

PDR Reference staff at 1 (800) 397–4209, 

(301) 415–4737, or by e-mail 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 petitioner/ 

requestor 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 petitioner/requestor 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 

petitioner/requestor who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party. 

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 petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor 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 


jlentini on DSKJ8SOYB1PROD with NOTICES
51338 Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 

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, 

which the NRC promulgated in August 

28, 2007, (72 FR 49139). 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 

a waiver in accordance with the 

procedures described below. 

To comply with the procedural 

requirements of E-Filing, at least five (5) 

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearing.docket@nrc.gov, or by calling 

(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/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

representative) already holds an NRCissued digital ID certificate). Each 

petitioner/requestor will need to 

download the Workplace Forms 

ViewerTM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

is available at http://www.nrc.gov/sitehelp/e-submittals/install-viewer.html. 

Information about applying for a digital 

ID certificate is available on NRC’s 

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

site-help/e-submittals/applycertificates.html. 

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it 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 filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

the EIE 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 

EIE 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 through the 

‘‘Contact Us’’ link located on the NRC 

Web site at http://www.nrc.gov/sitehelp/e-submittals.html or by calling the 

NRC Meta-System Help Desk, which is 

available between 8 a.m. and 8 p.m., 

Eastern Time, Monday through Friday, 

excluding government holidays. The 

Meta-System Help Desk can be 

contacted by telephone at 1–866–672– 

7640 or by e-mail at 

MSHD.Resource@nrc.gov. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file a 

motion, 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. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

the Atomic Safety and Licensing Board 

that the petition and/or request should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

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, an Atomic Safety and 

Licensing Board, or a 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. 

Pacific Gas and Electric Company, 

Docket No. 50–323, Diablo Canyon 

Nuclear Power Plant, Unit No. 2, San 

Luis Obispo County, California 

Date of application for amendment: 

September 3, 2009, as supplemented on 

September 8, 2009. 

Brief description of amendment: The 

amendment revised the Diablo Canyon 

Power Plant, Unit No. 2 Technical 

Specification (TS) 3.7.1, ‘‘Main Steam 

Safety Valves (MSSVs),’’ by increasing 

the Power Range Neutron Flux High 

setpoint in TS Table 3.7.1–1 from 87 

percent rated thermal power (RTP) to 

106 percent RTP. This will allow the 

unit to operate at full power with one 

main steam safety valve, MS–2–RV–224, 

inoperable for the remainder of Cycle 

15. 

Date of issuance: September 17, 2009. 

Effective date: As of its date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: 208. 

Facility Operating License No. DPR– 

82: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. A public 

notice of the proposed amendment was 

published in The Tribune newspaper, 

located in San Luis Obispo, California, 


jlentini on DSKJ8SOYB1PROD with NOTICES
Federal Register / Vol. 74, No. 192 / Tuesday, October 6, 2009 / Notices 51339 

on September 11 and 12, 2009. The 

notice provided an opportunity to 

submit comments on the NRC staff’s 

proposed NSHC determination. 

The supplemental letter dated 

September 8, 2009, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration determination as 

published in The Tribune. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, consideration of public 

comments, state consultation, and final 

NSHC determination are contained in a 

safety evaluation dated September 17, 

2009. 

Attorney for licensee: Jennifer Post, 

Esq., Pacific Gas and Electric Company, 

P.O. Box 7442, San Francisco, California 

94120. 

NRC Branch Chief: Michael T. 

Markley. 

Dated at Rockville, Maryland, this 25th day 

of September 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–23780 Filed 10–5–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2008–0361] 

Notice of Availability for Comment of 

Draft Standard Review Plan for 

Renewal of Independent Spent Fuel 

Storage Installation Licenses and Dry 

Cask Storage System Certificates of 

Compliance 

AGENCY: U.S. Nuclear Regulatory 

Commission. 

ACTION: Notice of availability and 

opportunity to provide comments. 

DATES: Comments must be provided by 

December 21, 2009. 

FOR FURTHER INFORMATION CONTACT: Ata 

Istar, Structural Mechanics and 

Materials Branch, Division of Spent 

Fuel Storage and Transportation 

Division, Office of Nuclear Material 

Safety and Safeguards, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20005–0001. Telephone: (301) 492– 

3409; fax number: (301) 492–3342; 

e-mail: ata.istar@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The Nuclear Regulatory Commission 

(NRC) has prepared a draft Standard 

Review Plan (SRP) NUREG–1927, 

entitled ‘‘Standard Review Plan for 

Renewal of Independent Spent Fuel 

Storage Installation Licenses and Dry 

Cask Storage System Certificate of 

Compliance.’’ This draft SRP would 

provide guidance to the NRC staff when 

reviewing Safety Analyses Reports 

submitted by applicants for renewals of 

specific Independent Spent Fuel Storage 

Installation licenses or dry cask storage 

system certificates of compliance under 

10 CFR part 72. This draft SRP is related 

to the proposed rule published in the 

Federal Register on September 15, 2009 

(74 FR 47126). The NRC is soliciting 

public comments on this draft SRP, 

which will be considered before the 

NRC issues the final version. 

II. Further Information 

Documents related to this action are 

available electronically at the NRC’s 

Electronic Reading Room at http:// 

www.nrc.gov/reading-rm/adams.html. 

From this site, you can access the NRC’s 

Agencywide Documents Access and 

Management System (ADAMS), which 

provides text and image files of NRC’s 

public documents. The ADAMS 

accession numbers for the documents 

related to this notice are provided in the 

following table. If you do not have 

access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC 

Public Document Room (PDR) reference 

staff at 1–800–397–4209, 301–415–4737, 

or by e-mail to ata.istar@nrc.gov. 

Interim staff guidance 

documents 

ADAMS accession 

No. 

Draft of SRP 

NUREG–1927.

ML092510340. 

These documents may also be viewed 

electronically on the public computers 

located at the NRC’s PDR, O–1 F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, MD 20852. The PDR 

reproduction contractor will copy 

documents for a fee. Comments and 

questions on this draft SRP should be 

directed to Ata Istar, Structural 

Mechanics and Materials Branch, 

Division of Spent Fuel Storage and 

Transportation, Office of Nuclear 

Materials Safety and Safeguards, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20005–0001 by 

December 21, 2009. Comments received 

after this date will be considered if it is 

practical to do so, but assurance of 

consideration cannot be given to 

comments received after this date. 

Comments can also be submitted by 

telephone, fax, or e-mail to the 

following: Telephone: (301) 492–3409; 

fax number: (301) 492–3342; e-mail: 

ata.istar@nrc.gov. 

Dated at Rockville, Maryland, this 29th day 

of September 2009. 

For the U.S. Nuclear Regulatory 

Commission. 

Christopher M. Regan, 

Chief, Structural Mechanics and Materials 

Branch, Division of Spent Fuel Storage and 

Transportation, Office of Nuclear Material 

Safety and Safeguards. 

[FR Doc. E9–24051 Filed 10–5–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 50–361 and 50–362; NRC– 

2009–0439] 

Southern California Edison Company; 

San Onofre Nuclear Generating 

Station, Unit 2 and Unit 3; 

Environmental Assessment and 

Finding of No Significant Impact 

The U.S. Nuclear Regulatory 

Commission (NRC) is considering 

issuance of a temporary exemption from 

Title 10 of the Code of Federal 

Regulations (10 CFR) Part 50, Section 

50.46 and 10 CFR 50, Appendix K, for 

Facility Operating License Nos. NPF–10 

and NPF–15, issued to Southern 

California Edison Company (SCE, the 

licensee), for operation of the San 

Onofre Nuclear Generating Station 

(SONGS), Unit 2 and Unit 3, 

respectively, located in San Diego 

County, California. Therefore, as 

required by 10 CFR 51.21, the NRC is 

issuing this environmental assessment 

and finding of no significant impact. 

Environmental Assessment 

Identification of the Proposed Action 

The requirements in 10 CFR 50.46 

specifically, and 10 CFR 50, Appendix 

K implicitly, refer to the use of Zircaloy 

or ZIRLO cladding. Therefore, a 

temporary exemption is required to use 

fuel rods clad with an advanced 

zirconium-based alloy that is not either 

Zircaloy or ZIRLO. Unlike the current 

fuel assemblies, the lead fuel assemblies 

(LFAs) manufactured by AREVA NP 

will contain M5 alloy cladding material. 

The licensee has requested a temporary 

exemption to allow the use of M5 alloy 

cladding. 

The temporary exemption would 

allow up to 16 LFAs manufactured by 

AREVA NP with M5 alloy cladding 

material to be inserted into the SONGS 

Unit 2 or Unit 3 reactor cores during the 


jlentini on DSKJ8SOYB1PROD with NOTICES