Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 25753 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2012–0096] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses and Combined Licenses 

Involving No Significant Hazards 

Considerations 

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 or combined 

license, as applicable, upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from April 5, 

2012 to April 18, 2012. The last 

biweekly notice was published on April 

17, 2012 (77 FR 22808). 

ADDRESSES: You may access information 

and comment submissions related to 

this document, which the NRC 

possesses and is publicly available, by 

searching on http://www.regulations.gov 

under Docket ID NRC–2012–0096. 

You may submit comments by the 

following methods: 

• Federal Rulemaking Web Site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2012–0096. Address 

questions about NRC dockets to Carol 

Gallagher; telephone: 301–492–3668; 

email: Carol.Gallagher@nrc.gov. 

• Mail comments to: Cindy Bladey, 

Chief, Rules, Announcements, and 

Directives Branch (RADB), Office of 

Administration, Mail Stop: TWB–05– 

B01M, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001. 

• Fax comments to: RADB at 301– 

492–3446. 

For additional direction on accessing 

information and submitting comments, 

see ‘‘Accessing Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

SUPPLEMENTARY INFORMATION: 

I. Accessing Information and 

Submitting Comments 

A. Accessing Information 

Please refer to Docket ID NRC–2012– 

0096 when contacting the NRC about 

the availability of information regarding 

this document. You may access 

information related to this document, 

which the NRC possesses and is 

publicly available, by the following 

methods: 

• Federal Rulemaking Web Site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2012–0096. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may access publicly 

available documents online in the NRC 

Library at http://www.nrc.gov/readingrm/adams.html. To begin the search, 

select ‘‘ADAMS Public Documents’’ and 

then select ‘‘Begin Web-based ADAMS 

Search.’’ For problems with ADAMS, 

please contact the NRC’s Public 

Document Room (PDR) reference staff at 

1–800–397–4209, 301–415–4737, or by 

email to pdr.resource@nrc.gov. 

Documents may be viewed in ADAMS 

by performing a search on the document 

date and docket number. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

B. Submitting Comments 

Please include Docket ID NRC–2012– 

0096 in the subject line of your 

comment submission, in order to ensure 

that the NRC is able to make your 

comment submission available to the 

public in this docket. 

The NRC cautions you not to include 

identifying or contact information in 

comment submissions that you do not 

want to be publicly disclosed. The NRC 

posts all comment submissions at 

http://www.regulations.gov as well as 

entering the comment submissions into 

ADAMS, and the NRC does not edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information in 

their comment submissions that they do 

not want to be publicly disclosed. Your 

request should state that the NRC will 

not edit comment submissions to 

remove such information before making 

the comment submissions available to 

the public or entering the comment 

submissions into ADAMS. 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses and Combined Licenses, 

Proposed No Significant Hazards 

Consideration Determination, and 

Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR) 50.92, this means 

that operation of the facility in 

accordance with the proposed 

amendment would not (1) involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; (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. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license or 

combined license. Requests for a 

mstockstill on DSK4VPTVN1PROD with NOTICES

25754 Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 

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 NRC’s PDR, located at 

One White Flint North, Room O1–F21, 

11555 Rockville Pike (first floor), 

Rockville, Maryland 20852. NRC 

regulations are accessible electronically 

from the NRC Library on the NRC’s Web 

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

doc-collections/cfr/. If a request for a 

hearing or petition for leave to intervene 

is filed by the above date, the 

Commission or a presiding officer 

designated by the Commission or by the 

Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, then any hearing held 

would take place before the issuance of 

any amendment. 

All documents filed in the NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

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

accordance with the NRC’s E-Filing rule 

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

submit and serve all adjudicatory 

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 10 

days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by email at 

hearing.docket@nrc.gov, or by telephone 

at 301–415–1677, to request (1) a digital 

identification (ID) certificate, which 

allows the participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

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

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on the 

NRC’s public Web site at http:// 

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

apply-certificates.html. System 

requirements for accessing the ESubmittal server are detailed in the 

NRC’s ‘‘Guidance for Electronic 

Submission,’’ which is available on the 

agency’s public Web site at http:// 

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

attempt to use other software not listed 

on the Web site, but should note that the 

NRC’s E-Filing system does not support 

unlisted software, and the NRC Meta 

System Help Desk will not be able to 

offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through the Electronic 

Information Exchange System, users 

will be required to install a Web 

browser plug-in from the NRC’s Web 

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

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with the NRC 

guidance available on the NRC’s public 

Web site at http://www.nrc.gov/sitehelp/e-submittals.html. A filing is 

considered complete at the time the 

documents are submitted through the 

NRC’s E-Filing system. To be timely, an 

electronic filing must be submitted to 

the E-Filing system no later than 11:59 

p.m. Eastern Time on the due date. 

Upon receipt of a transmission, the EFiling system time-stamps the document 

and sends the submitter an email notice 

confirming receipt of the document. The 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 25755 

E-Filing system also distributes an email 

notice that provides access to the 

document to the NRC’s Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the agency’s adjudicatory E-Filing 

system may seek assistance by 

contacting the NRC Meta System Help 

Desk through the ‘‘Contact Us’’ link 

located on the NRC’s Web site at 

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

MSHD.Resource@nrc.gov, or by a tollfree call at 1–866 672–7640. The NRC 

Meta System Help Desk is available 

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

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

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

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. A presiding 

officer, having granted an exemption 

request from using E-Filing, may require 

a participant or party to use E-Filing if 

the presiding officer subsequently 

determines that the reason for granting 

the exemption from use of E-Filing no 

longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in the NRC’s 

electronic hearing docket which is 

available to the public at http:// 

ehd1.nrc.gov/ehd/, unless excluded 

pursuant to an order of the Commission, 

or the presiding officer. Participants are 

requested not to include personal 

privacy information, such as social 

security numbers, home addresses, or 

home phone numbers in their filings, 

unless an NRC regulation or other law 

requires submission of such 

information. With respect to 

copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

Petitions for leave to intervene must 

be filed no later than 60 days from the 

date of publication of this notice. Nontimely filings will not be entertained 

absent a determination by the presiding 

officer that the petition or request 

should be granted or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

For further details with respect to this 

license amendment application, see the 

application for amendment which is 

available for public inspection at the 

NRC’s PDR, located at One White Flint 

North, Room O1–F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland 

20852. Publicly available documents 

created or received at the NRC are 

accessible electronically through 

ADAMS in the NRC Library at 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’s PDR 

Reference staff at 1–800–397–4209, 301– 

415–4737, or by email to 

pdr.resource@nrc.gov. 

Detroit Edison, Docket No. 50–341, 

Fermi 2, Monroe County, Michigan 

Date of amendment request: January 

10, 2012. 

Description of amendment request: 

The proposed amendment would 

modify Fermi 2 Plant Operating License, 

Appendix A, Technical Specifications 

(TS) to revise the Residual Heat 

Removal (RHR) Suppression Pool 

Cooling Surveillance Requirement (SR) 

3.6.2.3.2, flow requirement from greater 

than or equal to 10,000 gallons per 

minute (gpm) to greater than or equal to 

9,250 gpm. This change is consistent 

with the RHR suppression cooling rate 

associated with RHR heat exchanger 

minimum thermal performance 

requirements. Additionally, the 

proposed license amendment clarifies 

that SR 3.6.2.3.2 applies only to pumps 

required for meeting the Limiting 

Condition of Operation. 

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. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed TS SR 3.6.2.3.2 minimum 

flow of greater than or equal to 9,250 gpm is 

consistent with that assumed for accident 

extrapolation calculations of measured 

thermal performance obtained during RHR 

heat exchanger testing. This testing is 

performed to periodically demonstrate that 

the actual heat exchanger thermal 

performance exceeds that assumed for 

establishing the maximum post-accident bulk 

average suppression pool temperature. 

Therefore, the change in required RHR 

suppression pool cooling flow will not result 

in any increase in post-accident suppression 

pool temperature above that already 

evaluated for demonstrating adequate Net 

Pump Suction Head (NPSH) for any 

Emergency Core Cooling System (ECCS) 

pump. The change in the applicability of the 

surveillance to each required RHR pump 

provides consistency with the design of the 

system and maintains full capability of each 

RHR suppression pool cooling subsystem to 

provide post accident design basis cooling. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change revises TS SR 

3.6.2.3.2 for RHR suppression pool cooling 

flow to be consistent with that assumed for 

evaluating measured heat exchanger thermal 

performance against the minimum 

requirements of the plant safety analysis. 

Changing the applicability of the surveillance 

to each required RHR pump is consistent 

with the system design requirement and 

maintains full capability of each RHR 

suppression pool cooling subsystem to 

provide the post accident cooling function. 

No physical changes are being made to the 

installed RHR system or the manner in which 

it is operated. No new or different accident 

scenarios are created by this change. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The RHR system has historically been 

capable of meeting TS SR 3.6.2.3.2. This 

Surveillance requires demonstration of a 

system flow, in conjunction with a 

prescribed RHR heat exchanger capacity that 

ensures the overall suppression pool cooling 

capacity meets the requirements of the safety 

mstockstill on DSK4VPTVN1PROD with NOTICES

25756 Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 

analysis. However, the lack of available 

operating margin inherent in the design 

orifices of the RHR suppression pool cooling 

test return line and identification of a nonconservative bias in the test flow instrument 

calibration have eroded the flow test margin 

such that it is possible that the TS SR may 

not be satisfied in the future even though a 

large margin is maintained compared to the 

minimum performance assumed in the 

containment safety analyses. The proposed 

change makes the margin between TS SR 

3.6.2.3.2 and the performance assumed in the 

plant safety analyses available as a design 

and operating margin. This is ensured by 

establishing a higher level of required heat 

exchanger performance, where ample margin 

is available. Heat exchanger testing is 

conducted in accordance with existing 

testing standards as prescribed by EPRI TR– 

107397, Service Water Heat Exchanger 

Testing Guidelines. The minimum required 

flow rate necessary to satisfy RHR 

suppression pool cooling TS SR 3.6.2.3.2 will 

be documented in the plant design basis with 

the minimum required flow adjusted upward 

as necessary to account for instrument 

uncertainty and bias as well as differences 

between assumed accident and actual test 

operating conditions. 

The change in the applicability of the 

surveillance to each required RHR pump is 

consistent with the design basis of the plant 

and maintains full capability of the system to 

provide its safety related cooling function 

following a design basis 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: Bruce R. 

Masters, DTE Energy, General Council— 

Regulatory, 688 WCB, One Energy Plaza, 

Detroit, MI 48226–1279. 

NRC Acting Branch Chief: Shawn A. 

Williams. 

Detroit Edison, Docket No. 50–341, 

Fermi 2, Monroe County, Michigan 

Date of amendment request: January 

10, 2012. 

Description of amendment request: 

The proposed amendment would 

modify Fermi 2 Plant Operating License, 

Appendix A, Technical Specifications 

(TS) to modify Surveillance 

Requirement (SR) 3.4.3.2, in TS 3.4.3, 

‘‘Safety Relief Valves (SRVs)’’, SR 

3.5.1.13, in TS 3.5.1, ‘‘ECCS-Operating,’’ 

and SR 3.6.1.6.1, in TS 3.6.1.6, ‘‘LowLow Set (LLS) Valves.’’ This proposed 

amendment replaces the current 

requirement in these TS SRs to verify 

the SRV opens when manually actuated 

with an alternate requirement that 

verifies the SRV is capable of being 

opened. The verification of that 

capability would be satisfied by a series 

of overlapping tests, performed during a 

refueling outage, that demonstrate the 

required functions of successive valve 

stages. 

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. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed change does not modify the 

method of demonstrating the operability of 

the Safety Relief Valves (SRVs) in both the 

safety and relief modes of operation. As 

currently stated in the Technical 

Specification (TS) Bases ‘‘* * * valve 

OPERABILITY and the setpoints for 

overpressure protection are verified, per 

ASME Code requirements, prior to valve 

installation.’’ The proposed change does 

modify the method for demonstrating the 

proper mechanical functioning of the SRVs. 

The SRVs are required to function in the 

safety mode to prevent overpressurization of 

the reactor vessel and reactor coolant system 

pressure boundary during various analyzed 

transients, including Main Steam Isolation 

Valve closure. SRVs associated with the 

Automatic Depressurization System are also 

required to function in the relief mode to 

reduce reactor pressure to permit injection by 

low pressure Emergency Core Cooling 

System (ECCS) pumps during certain reactor 

coolant pipe break accidents. The current 

testing method demonstrates the proper 

mechanical functioning of the SRVs in both 

modes through manual actuation of the 

SRVs. The proposed new testing method 

demonstrates both operability and proper 

mechanical functioning using a series of 

overlapping tests that demonstrate proper 

functioning of the SRV and supporting 

control components. This proposed testing 

method results in acceptable demonstration 

of the SRV functions in both the safety and 

relief modes, and therefore provides 

assurance that the probability of SRV failure 

will not increase. None of the accident safety 

analyses is affected by the requested TS 

changes. Therefore, the consequences of 

accidents mitigated by the SRVs will not 

increase. 

Certain SRV malfunctions are included in 

the UFSAR safety analyses. Specifically, the 

plant safety analyses include the inadvertent 

opening of an SRV and a stuck open SRV. By 

reducing or not actuating the SRVs during 

plant operation for testing and thus reducing 

the potential incidence of pilot stage leakage 

of the SRVs, the proposed testing reduces a 

contributor to these events. 

Based on these considerations, the 

proposed test method does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change modifies the method 

of testing of the SRVs, but does not alter the 

functions or functional capabilities of the 

SRVs. Testing under the proposed method is 

performed in offsite test facilities and in the 

plant during outage periods when the SRV 

functions are not required. Existing analyses 

address events involving an SRV 

inadvertently opening or failing to reclose. 

Analyses also address the failure of one or 

more SRVs to open. The proposed change 

does not introduce any new failure mode, 

and therefore, does not create the possibility 

of a new or different kind of accident from 

any accident previously evaluated. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The proposed amendment provides for an 

alternative means of testing the SRVs. The 

proposed changes will provide a complete 

verification of the functional capability of the 

SRVs by performing a series of tests, 

inspections, and maintenance activities 

without opening the valves with reactor 

steam while installed in the plant. The 

alternative testing and associated 

programmatic controls will provide an 

equivalent level of assurance that the SRVs 

are capable of performing their intended 

accident mitigation safety functions. The 

proposed amendment does not affect the 

valve setpoints or adversely affect any other 

operational criteria assumed for accident 

mitigation. No changes are proposed that 

alter the setpoints at which protective actions 

are initiated, and there is no change to the 

operability requirements for equipment 

assumed to operate for accident mitigation. 

Moreover, it is expected that the alternative 

testing methodology will increase the margin 

of safety by reducing the potential for SRV 

leakage resulting from testing the SRVs with 

reactor steam pressure while installed in the 

plant. 

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: Bruce R. 

Masters, DTE Energy, General Council— 

Regulatory, 688 WCB, One Energy Plaza, 

Detroit, MI 48226–1279. 

NRC Acting Branch Chief: Shawn A. 

Williams. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 25757 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: 

September 20, 2011, as supplemented 

by letter dated November 21, 2011. 

Description of amendment request: 

The proposed amendments would allow 

revisions to the current licensing basis 

to allow a measurement uncertainty 

recapture (MUR) power uprate. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment changes the 

rated thermal power from 2568 megawatts 

thermal (MWt) to 2610 MWt; an increase of 

approximately 1.64% Rated Thermal Power. 

Duke Energy’s evaluations have shown that 

all structures, systems and components 

(SSCs) are capable of performing their design 

function at the uprated power of 2610 MWt. 

A review of station accident analyses found 

that all but two analyses remain bounding at 

the uprated power of 2610 MWt. These two 

analyses (High Energy Line Break and Double 

Main Steam Line Break) were reanalyzed at 

the higher power level and found to be 

acceptable. 

The radiological consequences of operation 

at the uprated power conditions have been 

assessed. The proposed power uprate does 

not affect release paths, frequency of release, 

or the analyzed reactor core fission product 

inventory for any accidents previously 

evaluated in the Final Safety Analysis 

Report. Analyses performed to assess the 

effects of mass and energy releases remain 

valid. All acceptance criteria for radiological 

consequences continue to be met at the 

uprated power level. 

As summarized in Sections IV, V, and VI 

of Enclosure 2, the proposed change does not 

involve any change to the design or 

functional requirements of the associated 

systems. That is, the increased power level 

neither degrades the performance of, nor 

increases the challenges to any safety systems 

assumed to function in the plant safety 

analysis. 

While power level is an input to accident 

analyses, it is not an initiator of accidents. 

The proposed change does not affect any 

accident precursors and does not introduce 

any accident initiators. The proposed change 

does not impact the usefulness of the 

Surveillance Requirements (SRs) in 

evaluating the operability of required systems 

and components. 

In addition, evaluation of the proposed TS 

[Technical Specification] change 

demonstrates that the availability of 

equipment and systems required to prevent 

or mitigate the radiological consequences of 

an accident is not significantly affected. 

Since the impact on the systems is minimal, 

it is concluded that the overall impact on the 

plant safety analysis is negligible. 

Therefore, the proposed TS changes do not 

significantly increase the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

A Failure Modes and Effects Analysis of 

the new system was performed, and the 

possible effects of failures of the new 

equipment and the increased power level on 

the overall plant systems were reviewed. 

This review found that no new or different 

accidents were created by the new equipment 

or the uprated power levels. 

No installed equipment is being operated 

in a different manner. The proposed changes 

have no significant adverse affect on any 

safety-related SSCs and do not significantly 

change the performance or integrity of any 

safety-related system. 

The proposed changes do not adversely 

affect any current system interfaces or create 

any new interfaces that could result in an 

accident or malfunction of a different kind 

than previously evaluated. The uprated 

power does not create any new accident 

initiators. Credible malfunctions are bounded 

by the current accident analyses of record or 

recent evaluations demonstrating that 

applicable criteria are still met with the 

proposed changes. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Although the proposed amendment 

increases the operating power level of the 

plants, it retains the margin of safety because 

it is only increasing power by the amount 

equal to the reduction in uncertainty in the 

heat balance calculation. The margins of 

safety associated with the power uprate are 

those pertaining to core thermal power. 

These include fuel cladding, reactor coolant 

system pressure boundary, and containment 

barriers. Analyses demonstrate that the 

current design basis continues to be met after 

the MUR power uprate. Components 

associated with the reactor coolant system 

pressure boundary structural integrity, 

including pressure-temperature limits, vessel 

fluence, and pressurized thermal shock are 

bounded by the current analyses. Systems 

will continue to operate within their design 

parameters and remain capable of performing 

their intended safety functions. 

The current Oconee safety analyses, and 

the revised design basis radiological accident 

dose calculations, bound the power uprate 

and therefore do not significantly impact 

margins. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lara S. Nichols, 

Associate General Counsel, Duke Energy 

Corporation, 526 South Church Street— 

EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: 

December 8, 2011. 

Description of amendment request: 

The amendment would revise Technical 

Specifications (TS) 3.8.1; ‘‘AC 

[Alternating Current] Sources— 

Operating.’’ Specifically, the 

amendment would revise TS 3.8.1 and 

the associated Bases, to expand its scope 

to include provisions for testing of the 

automatic transfer function from the 

station 22 kiloVolt (kV) bus to offsite 

power for Division III. A new 

Surveillance Requirement (SR) would 

be added to ensure availability of offsite 

power after loss of the station (onsite) 22 

kV bus when offsite power remains 

available. The amendment would also 

add notes to the Limiting Condition for 

Operation (LCO) and SR to require this 

feature when Division III is powered by 

onsite power. In addition, new 

ACTIONS would be added to ensure 

this transfer from onsite to offsite is 

maintained when a required offsite 

power source is lost. 

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 the Technical 

Specification Surveillance Requirements to 

allow power for emergency systems to be 

supplied from onsite power prior to event 

initiation. This power supply will be 

transferred to the current accepted offsite 

power source if the main generator is no 

longer available. The proposed Surveillance 

Requirement is to confirm the automatic 

transfer function. 

The proposed changes do not involve a 

change in the design requirements of the 

electrical power systems, including the 

emergency power systems. The plant will 

continue to operate within acceptable 

parameters (electrical loading, etc.) The 

mstockstill on DSK4VPTVN1PROD with NOTICES

25758 Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 

proposed changes do not change the function 

of plant equipment, or affect the response of 

emergency power systems. 

The proposed changes do not involve a 

change in the design basis initiators for loss 

of offsite power to the emergency power 

systems. The proposed change utilizes 

existing components and circuits. The 

change will add a new surveillance 

requirement to confirm the design function 

operation. 

The proposed change does not impact 

other design basis accident initiators or 

analyzed events or assumed mitigation of 

accident or transient events. 

The proposed change does not involve a 

change to the consequences of a design basis 

event as described in the Safety Analysis 

Report. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises the Technical 

Specification surveillance requirements to 

confirm operation of existing components 

and circuits. The proposed changes do not 

involve a change in the design basis initiators 

for loss of offsite power to the emergency 

power systems. 

The proposed changes do not involve a 

change in the design requirements of the 

electrical power systems, including the 

emergency power systems. The proposed 

changes do not change the function of plant 

equipment, nor do they affect the response of 

emergency power systems. 

The proposed changes do not involve a 

change in the operational limits or physical 

design of the electrical power systems, 

particularly the emergency power systems. 

The proposed changes do not change the 

design function or operation of plant 

equipment, nor do they introduce any new 

failure mechanisms. This change will 

implement surveillance requirements to 

confirm the design function operation. 

The transfer function components 

supporting the safety-related buses have been 

designed to applicable quality standards and 

design criteria. As such, no new failure 

modes are being introduced. The plant 

equipment will continue to respond in 

accordance with the design and analyses, and 

no malfunction of a new or different type is 

being introduced by the proposed changes. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change revises the Technical 

Specification surveillance requirements. The 

proposed changes do not involve a change in 

the operational limits or the response of that 

equipment if it is called upon to operate. 

The performance capability of the 

emergency diesel generators will not be 

affected. The plant will continue to operate 

within acceptable parameters (electrical 

loading, etc.)[.] 

In addition, administrative controls will 

ensure there are adequate administrative 

controls are in place to ensure the plant 

configuration remains as evaluated. 

The results of the PRA performed to 

quantitatively assess the risk impact of this 

change indicate there is a minimal risk 

impact. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit 2, Westchester 

County, New York 

Date of amendment request: January 

11, 2012. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

Table 3.3.6–1, ‘‘Containment Purge 

System and Pressure Relief Line 

Isolation Instrumentation.’’ The 

proposed amendment would change the 

term ‘‘ALLOWABLE VALUE’’ to ‘‘TRIP 

SETPOINT’’ and revise the current 

setpoint used for the Containment Purge 

Systems and Pressure Relief Line 

isolation. The proposed revision to TS 

Table 3.3.6–1 will change ‘‘≤ 3 x 

background’’ to allow the trip setpoint 

to be as specified in the Offsite Dose 

Calculation Manual (ODCM). 

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 will revise the term 

‘‘ALLOWABLE VALUE’’ to ‘‘TRIP 

SETPOINT’’ and change the setpoint 

requirements from ‘‘≤ 3 [x] background’’ to 

allow the allowable value to be as specified 

in the Offsite Dose Calculation Manual 

(ODCM). The change to trip setpoint is a 

correction of an administrative error and will 

only affect the instrument setting specified. 

Therefore it does not involve the initiation of 

an accident or the consequences. The values 

for the instrument setting are provided for 

isolating the Containment Purge and Pressure 

Relief Systems due to increased source terms 

and are redundant to containment isolation 

signals. They have no effect on the 

probability of an accident previously 

evaluated. The change in the setting will be 

negligible for purposes of an accident 

termination. The ODCM limits are based on 

10 CFR 20 [Title 10 of the Code of Federal 

Regulations, Part 20] limits which are 

substantially below accident analysis release 

rates. Therefore the change has a minimum 

effect on the consequences of such accidents. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of previously 

evaluated accidents. 

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 will revise the term 

‘‘ALLOWABLE VALUE’’ to ‘‘TRIP 

SETPOINT’’ and change the setpoint 

requirements. The changes do not affect the 

system operations, plant operating 

procedures or affect how the plant is 

operated. The change does not create the 

possibility of any equipment failure or effect 

on other equipment. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change will revise the term 

‘‘ALLOWABLE VALUE’’ to ‘‘TRIP 

SETPOINT’’ and change the setpoint 

requirements. The change to trip setpoint is 

correcting an administrative error and has no 

significant affect on the margin of safety. The 

proposed change involves changes to existing 

setpoints for automatic isolation of the 

Containment Purge and Pressure Relief 

Systems. However, the ability of the systems 

to isolate remains within current evaluations 

and therefore does not significantly reduce 

the safety margin. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue White Plains, NY 

10601. 

NRC Branch Chief: George Wilson. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 25759 

Northern States Power Company— 

Minnesota, Docket No. 50–263, 

Monticello Nuclear Generating Plant 

(MNGP), Wright County, Minnesota 

Date of amendment request: April 5, 

2012. 

Description of amendment request: 

The licensee proposed to revise the 

MNGP Technical Specifications (TS) 

3.3.5.1, ‘‘Emergency Core Cooling 

System (ECCS) Instrumentation.’’ 

Specifically, it is proposed to revise the 

lower allowable value limit for Table 

3.3.5.1, Functions 1.e and 2.e, ‘‘Reactor 

Steam Dome Pressure Permissive— 

Bypass Timer (Pump Permissive).’’ The 

licensee has determined that the upper 

allowable value limit for the Automatic 

Depressurization System (ADS) bypass 

timer function provides the operator 

sufficient time to assess the situation 

and inhibit ADS actuation if the event 

does not require rapid reactor 

depressurization. The lower allowable 

value limit ADS bypass timer function 

pertains to providing adequate margin 

to unwanted pump starts during reactor 

water level transients and is not 

credited in the safety analyses. 

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 (NSHC). The NRC staff 

reviewed the licensee’s NSHC analysis 

and prepared its own as follows: 

(1) Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change does not physically 

impact the plant nor does it impact any 

design or functional requirements of the 

Automatic Depressurization System (ADS). 

The proposed change does not degrade the 

performance or increase the challenges to any 

safety systems assumed to function in the 

accident analysis. There is no change to 

normal plant operating parameters or 

accident mitigation performance. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

(2) Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

There are no hardware changes nor are 

there any changes in the method by which 

plant systems perform a safety function. This 

request does not affect the normal method of 

plant operation. No new equipment is 

introduced which could create a new or 

different kind of accident. No new equipment 

failure modes are created. No new accident 

scenarios, transient precursors, failure 

mechanisms, or limiting single failures are 

introduced. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

(3) Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment does not affect 

the assumptions of the safety analysis or the 

availability or operability of any plant 

equipment. There is no reduction in the 

margin of safety because the criteria for the 

performance of the ADS are not changed and 

there are no changes to those plant systems 

necessary to assure the accomplishment of 

protection functions. 

Therefore, the proposed amendment does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on its 

own analysis, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the proposed 

amendment involves no significant 

hazards consideration. 

Attorney for the licensee: Peter M. 

Glass, Assistant General Counsel, Xcel 

Energy Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Istvan Frankl, 

Acting. 

Notice of Issuance of Amendments to 

Facility Operating Licenses and 

Combined Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR chapter I, which are set forth in 

the license amendment. 

A notice of consideration of issuance 

of amendment to facility operating 

license or combined license, as 

applicable, proposed no significant 

hazards consideration determination, 

and opportunity for a hearing in 

connection with these actions, was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the NRC’s Public Document Room 

(PDR), located at One White Flint North, 

Room O1–F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland 20852. 

Publicly available documents created or 

received at the NRC are accessible 

electronically through the Agencywide 

Documents Access and Management 

System (ADAMS) in the NRC Library at 

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’s Reference 

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

or by email to pdr.resource@nrc.gov. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station 

(DBNPS), Unit 1, Ottawa County, Ohio 

Date of application for amendment: 

May 20, 2011 as supplemented by letter 

dated February 7, 2012. 

Brief description of amendment: This 

amendment revised Technical 

Specification (TS) 5.5.8.g to perform the 

special visual inspections based on a 

condition rather than a specific 

frequency. Specifically, TS 5.5.8.g 

requires visual inspection of the secured 

internal auxiliary feedwater header 

(AFWH), header to shroud attachment 

welds, and external header thermal 

sleeves of the steam generators (SGs) at 

DBNPS to be performed during the third 

period of each 10-year inservice 

inspection interval (ISI). With the 

proposed change, if eddy current 

inspections (required by TS 5.5.8.d.5) 

identify any SG peripheral rube to 

secured internal AFWH gaps less than @ 

inches or there is evidence that the 

header is degrading or has moved, then 

the TS 5.5.8.g visual inspections shall 

be performed on the affected SG. 

Date of issuance: April 18, 2012. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 285. 

Facility Operating License No. NPF–3: 

Amendment revised the Technical 

Specifications and License. 

Date of initial notice in Federal 

Register: September 20, 2011 (76 FR 

58306). The February 7, 2012 

mstockstill on DSK4VPTVN1PROD with NOTICES

25760 Federal Register / Vol. 77, No. 84 / Tuesday, May 1, 2012 / Notices 

supplement 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 finding of no significant 

hazards consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 18, 2012. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–410, Nine Mile Point 

Nuclear Station (NMP2), Unit 2, Oswego 

County, New York 

Date of application for amendment: 

December 30, 2011, as supplemented on 

March 20, 2012. 

Brief description of amendment: The 

proposed amendment changes the 

NMP2 Updated Safety Analysis Report 

allowing the use of Modified Alloy 718 

material for fabrication of the NMP2 

reactor recirculation system jet pump 

holddown beams. 

Date of issuance: April 13, 2012. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment No.: 141. 

Renewed Facility Operating License 

No. NPF–069: The amendment revises 

the License and Updated Safety 

Analysis Report. 

Date of initial notice in Federal 

Register: February 8, 2012 (77 FR 

6601). The supplemental letter dated 

March 20, 2012, provided additional 

information that clarified the 

application and did not expand the 

scope of the application as originally 

noticed, and did not change the Nuclear 

Regulatory Commission staff’s initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 13, 2012. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, et 

al., Docket Nos. 50–280 and 50–281, 

Surry Power Station, Units 1 and 2, 

Surry County, Virginia 

Date of application for amendments: 

July 28, 2011, as supplemented on 

February 14 and March 14, 2012. 

Brief Description of amendments: 

These amendments permanently revise 

the Technical Specifications (TS) 6.4.Q, 

‘‘Steam Generator (SG) Program,’’ to 

exclude portions of the SG tube below 

the top of the SG tubesheet from 

periodic inspections. In addition, this 

amendment request proposes to revise 

TS 6.6.A.3, ‘‘Steam Generator Tube 

Inspection Report,’’ to remove 

references to the previous Unit 1 onetime and Unit 2 temporary alternate 

repair criteria and provides reporting 

requirements specific to the permanent 

alternate repair criteria. This 

amendment also addressed minor 

administrative revisions to reinstate the 

superscript number 1 as the end of the 

TS 4.13.B. 

Date of issuance: April 17, 2012. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: Unit 1—277 and 

Unit 2—277. 

Renewed Facility Operating License 

Nos. DPR–32 and DPR–37: Amendments 

change the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: October 25, 2011 (76 FR 

66090). The supplements dated 

February 14, 2012 and March 14, 2012, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of the 

amendments is contained in a Safety 

Evaluation dated April 17, 2012. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 20th day 

of April 2012. 

For the Nuclear Regulatory Commission. 

Michele G. Evans, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2012–10195 Filed 4–30–12; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2011–0183] 

Low-Level Radioactive Waste 

Management and Volume Reduction 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Policy statement; issuance. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC or the Commission) 

is revising its 1981 Policy Statement on 

Low-Level Radioactive Waste (LLRW) 

Volume Reduction (Policy Statement). 

This statement encouraged licensees to 

take steps to reduce the amount of waste 

generated and to reduce the volume of 

waste once generated. The purpose of 

this revised statement is to recognize 

that progress in reducing waste volume 

has been achieved since the 1981 Policy 

Statement was published, and to 

acknowledge that factors other than 

volume reduction may be considered by 

licensees to determine how best to 

manage their LLRW. 

DATES: This Policy Statement is effective 

on May 1, 2012. 

ADDRESSES: Please refer to Docket ID 

NRC–2011–0183 when contacting the 

NRC about the availability of 

information for this policy statement. 

You may access information and 

comment submissions related to this 

policy statement, which the NRC 

possesses and are publicly available, by 

the following methods: 

• Federal Rulemaking Web site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2011–0183. Address 

questions about NRC dockets to Carol 

Gallagher; telephone: 301–492–3668; 

email: Carol.Gallagher@nrc.gov. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may access publicly 

available documents online in the NRC 

Library at http://www.nrc.gov/readingrm/adams.html. To begin the search, 

select ‘‘ADAMS Public Documents’’ and 

then select ‘‘Begin Web-based ADAMS 

Search.’’ For problems with ADAMS, 

please contact the NRC’s Public 

Document Room (PDR) reference staff at 

1–800–397–4209, 301–415–4737, or by 

email to pdr.resource@nrc.gov. The 

ADAMS accession number for each 

document referenced in this notice (if 

that document is available in ADAMS) 

is provided the first time that a 

document is referenced. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

FOR FURTHER INFORMATION CONTACT: 

Donald Lowman, Office of Federal and 

State Materials and Environmental 

Management Programs, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone: 301–415– 

5452, email: Donald.Lowman@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Background 

In 1981, the NRC published a Policy 

Statement (46 FR 51100; October 16, 

1981) regarding the volume reduction of 

LLRW. The Policy Statement addressed: 

• The need for a volume reduction 

policy; and 

• The need for waste generators to 

minimize the quantity of waste 

produced. 

For 30 years, this Policy Statement 

has conveyed the Commission’s 

expectations that generators of LLRW 

mstockstill on DSK4VPTVN1PROD with NOTICES