Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 21917 

Drug Schedule 

Methadone (9250) ........................ II 

Methadone intermediate (9254) ... II 

Dextropropoxyphene, bulk (nondosage forms) (9273).II 

Morphine (9300) ........................... II 

Oxymorphone (9652) ................... II 

Oxycodone (9143) ........................ II 

The company plans to manufacture 

the above listed controlled substances in 

bulk for distribution to its customers. 

Drug codes 1100 (amphetamine) and 

2550 (glutethimide) have been 

withdrawn from the application for 

registration at the request of the 

company. 

No comments or objections have been 

received. DEA has considered the 

factors in 21 U.S.C. 823(a) and 

determined that the registration of 

Siegfried (USA), Inc. to manufacture the 

listed basic classes of controlled 

substances is consistent with the public 

interest at this time. DEA has 

investigated Siegfried (USA), Inc. to 

ensure that the company’s registration is 

consistent with the public interest. The 

investigation has included inspection 

and testing of the company’s physical 

security systems, verification of the 

company’s compliance with State and 

local laws, and a review of the 

company’s background and history. 

Therefore, pursuant to 21 U.S.C. 823(a), 

and in accordance with 21 CFR 1301.33, 

the above named company is granted 

registration as a bulk manufacturer of 

the basic classes of controlled 

substances listed. 

Dated: April 11, 2011. 

Joseph T. Rannazzisi, 

Deputy Assistant Administrator, Office of 

Diversion Control, Drug Enforcement 

Administration. 

[FR Doc. 2011–9361 Filed 4–18–11; 8:45 am] 

BILLING CODE 4410–09–P 

DEPARTMENT OF JUSTICE 

Drug Enforcement Administration 

Manufacturer of Controlled 

Substances; Notice of Registration 

By Notice dated October 8, 2010, and 

published in the Federal Register on 

October 20, 2010, (75 FR 64744), 

Cayman Chemical Company, 1180 East 

Ellsworth Road, Ann Arbor, Michigan 

48108, made application by renewal to 

the Drug Enforcement Administration 

(DEA) to be registered as a bulk 

manufacturer of the following basic 

classes of controlled substances: 

Drug Schedule 

Marihuana (7360) ......................... I 

Tetrahydrocannabinols (7370) ..... I 

The company plans to manufacture 

small quantities of marihuana 

derivatives for research purposes. In 

reference to drug code 7360 

(Marihuana), the company plans to bulk 

manufacture cannabidiol. In reference to 

drug code 7370 

(Tetrahydrocannabinols), the company 

will manufacture a synthetic THC. No 

other activity for this drug code is 

authorized for this registration. 

No comments or objections have been 

received. DEA has considered the 

factors in 21 U.S.C. 823(a) and 

determined that the registration of 

Cayman Chemical Company to 

manufacture the listed basic classes of 

controlled substances is consistent with 

the public interest at this time. DEA has 

investigated Cayman Chemical 

Company to ensure that the company’s 

registration is consistent with the public 

interest. The investigation has included 

inspection and testing of the company’s 

physical security systems, verification 

of the company’s compliance with state 

and local laws, and a review of the 

company’s background and history. 

Therefore, pursuant to 21 U.S.C. 823, 

and in accordance with 21 CFR 1301.33, 

the above named company is granted 

registration as a bulk manufacturer of 

the basic classes of controlled 

substances listed. 

Dated: April 11, 2011. 

Joseph T. Rannazzisi, 

Deputy Assistant Administrator, Office of 

Diversion Control, Drug Enforcement 

Administration. 

[FR Doc. 2011–9360 Filed 4–18–11; 8:45 am] 

BILLING CODE 4410–09–P 

NATIONAL TRANSPORTATION 

SAFETY BOARD 

Sunshine Act Meeting 

TIME AND DATE: 9:30 a.m., Tuesday, April 

26, 2011. 

PLACE: NTSB Conference Center, 429 

L’Enfant Plaza, SW., Washington, DC 

20594. 

STATUS: The ONE item is open to the 

public. 

MATTER TO BE CONSIDERED: 

8093A Aviation Accident Report 

Crash During Unstabilized Approach, 

Empire Airlines Flight 8284, Avions de 

Transport Re ́gional Aerospatiale Alenia 

ATR 42 320, N902FX, Lubbock, Texas, 

January 27, 2009. 

NEWS MEDIA CONTACT: Telephone: (202) 

314–6100. 

The press and public may enter the 

NTSB Conference Center one hour prior 

to the meeting for set up and seating. 

Individuals requesting specific 

accommodations should contact 

Rochelle Hall at (202) 314–6305 by 

Friday, April 22, 2011. 

The public may view the meeting via 

a live or archived webcast by accessing 

a link under ‘‘News & Events’’ on the 

NTSB home page at http:// 

www.ntsb.gov. 

FOR FURTHER INFORMATION CONTACT: 

Candi Bing, (202) 314–6403 or by e-mail 

at bingc@ntsb.gov. 

Dated: April 15, 2011. 

Candi R. Bing, 

Federal Register Liaison Officer. 

[FR Doc. 2011–9565 Filed 4–15–11; 4:15 pm] 

BILLING CODE 7533–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2011–0082] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to Section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC) 

is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from March 24, 

2011, to April 6, 2011. The last biweekly 

notice was published on April 5, 2011 

(76 FR 18801). 

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. 

mstockstill on DSKH9S0YB1PROD with NOTICES

21918 Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), § 50.92, this 

means that operation of the facility in 

accordance with the proposed 

amendment would not (1) involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules, 

Announcements and Directives Branch 

(RADB), TWB–05–B01M, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be faxed to the RADB at 301–492– 

3446. Documents may be examined, 

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

Public Document Room (PDR), located 

at One White Flint North, Room O1– 

F21, 11555 Rockville Pike (first floor), 

Rockville, Maryland 20852. 

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, Room 

O1–F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland 20852. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, then any hearing held 

would take place before the issuance of 

any amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

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

accordance with the NRC E-Filing rule 

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

submit and serve all adjudicatory 

documents over the internet, or in some 

cases to mail copies on electronic 

storage media. Participants may not 

submit paper copies of their filings 

unless they seek an exemption in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

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Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 21919 

(10) days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by e-mail at 

hearing.docket@nrc.gov, or by telephone 

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

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 

NRC’s public Web site at http:// 

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

apply-certificates.html. System 

requirements for accessing the E– 

Submittal server are detailed in NRC’s 

‘‘Guidance for Electronic Submission,’’ 

which is available on the agency’s 

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

site-help/e-submittals.html. Participants 

may attempt to use other software not 

listed on the Web site, but should note 

that the NRC’s E-Filing system does not 

support unlisted software, and the NRC 

Meta System Help Desk will not be able 

to offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through the Electronic 

Information Exchange System, users 

will be required to install a Web 

browser plug-in from the NRC Web site. 

Further information on the Web-based 

submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC public Web site at 

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

complete at the time the documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

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

Time on the due date. Upon receipt of 

a transmission, the E-Filing system 

time-stamps the document and sends 

the submitter an e-mail notice 

confirming receipt of the document. The 

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

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the agency’s adjudicatory E-Filing 

system may seek assistance by 

contacting the NRC Meta System Help 

Desk through the ‘‘Contact Us’’ link 

located on the NRC Web site at http:// 

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

MSHD.Resource@nrc.gov, or by a tollfree call at 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 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 

Commission’s PDR, located at One 

White Flint North, Room O1–F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland 20852. Publicly available 

records will be accessible from the 

ADAMS Public Electronic Reading 

Room on the Internet at the NRC Web 

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

adams.html. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

Calvert Cliffs Nuclear Power Plant, LLC, 

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

Cliffs Nuclear Power Plant, Unit 1 and 

2, Calvert County, Maryland 

Date of amendments request: October 

25, 2010. 

Description of amendments request: 

The amendment would revise Technical 

Specification Limiting Condition for 

Operation (LCO) 3.0.5 to provide 

clarification as to when the LCO can be 

invoked in order to perform required 

testing to demonstrate OPERABILITY of 

equipment. 

mstockstill on DSKH9S0YB1PROD with NOTICES

21920 Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 

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 LCO 3.0.5 more 

clearly specifies the situations when LCO 

3.0.5 can be applied. In some Technical 

Specifications, the steps taken to comply 

with ACTIONS involve the placement of 

redundant or alternate equipment or trains 

into service, or the repositioning (e.g., 

opening or closing) or components. The 

proposed change would allow the use of LCO 

3.0.5 in situations such as these. This 

proposed change does not, however, change 

the intent of LCO 3.0.5. The purpose of LCO 

3.0.5 remains to provide an exception to LCO 

3.0.2, to not comply with the applicable 

Required Action(s) while performing 

required testing to demonstrate the 

OPERABILITY of either equipment being 

returned to service or the OPERABILITY of 

other equipment. 

The proposed change does not affect any 

analyzed accident initiators, nor does it 

change the units’ ability to successfully 

respond to any previously evaluated 

accident. As a result, there is also no change 

to existing radiological assumptions used in 

the accident evaluations. In addition this 

proposed change does not change the 

operation or maintenance performed on 

operating equipment. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to LCO 3.0.5 more 

clearly specifies the situations when LCO 

3.0.5 can be applied. In some Technical 

Specifications, the steps taken to comply 

with ACTIONS involve the placement of 

redundant or alternate equipment or trains 

into service, or the repositioning (e.g., 

opening or closing) or components. The 

proposed change would allow the use of LCO 

3.0.5 in situations such as these. This 

proposed change does not, however, change 

the intent of LCO 3.0.5. The purpose of LCO 

3.0.5 remains to provide an exception to LCO 

3.0.2, to not comply with the applicable 

Required Action(s) while performing 

required testing to demonstrate the 

OPERABILITY of either equipment being 

returned to service or the OPERABILITY of 

other equipment. 

The proposed change does not involve a 

modification to the physical configuration of 

the units nor does it involve any change in 

the methods governing normal plant 

operation. The proposed change does not 

impose any new or different requirements or 

introduce a new accident initiator, accident 

precursor, or malfunction mechanism. 

Additionally there is no change in the types 

or increase in the amounts of any effluent 

that may be released offsite and there is no 

increase in individual or cumulative 

occupational exposure. 

Therefore the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change to LCO 3.0.5 more 

clearly specifies the situations when LCO 

3.0.5 can be applied. In some Technical 

Specifications, the steps taken to comply 

with ACTIONS involve the placement of 

redundant or alternate equipment or trains 

into service, or the repositioning (e.g., 

opening or closing) or components. The 

proposed change would allow the use of LCO 

3.0.5 in situations such as these. This 

proposed change does not, however, change 

the intent of LCO 3.0.5. The purpose of LCO 

3.0.5 remains to provide an exception to LCO 

3.0.2, to not comply with the applicable 

Required Action(s) while performing 

required testing to demonstrate the 

OPERABILITY of either equipment being 

returned to service or the OPERABILITY of 

other equipment. 

The proposed change does not involve any 

modification to the physical configuration of 

the operating units and does not alter 

equipment operation. As such, the safety 

functions of plant equipment and their 

response to any analyzed accident scenario 

are unaffected by this proposed change and 

thus there is no reduction in the margin of 

safety. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety for the operation of each unit. 

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 

amendments request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Generation Group, LLC, 

750 East Pratt Street, 17th floor, 

Baltimore, MD 21202. 

NRC Branch Chief: Nancy L. Salgado. 

Calvert Cliffs Nuclear Power Plant, LLC, 

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

Cliffs Nuclear Power Plant, Unit 1 and 

2, Calvert County, Maryland 

Date of amendments request: March 

22, 2011. 

Description of amendments request: 

The proposed amendment would revise 

the Technical Specifications (TSs) to 

define a new time limit for restoring 

inoperable reactor coolant system (RCS) 

leakage detection instrumentation to 

operable status. The proposed TS 

changes are consistent with TS Task 

Force (TSTF)–513, ‘‘Revise PWR 

[pressurized-water reactor] Operability 

Requirements and Actions for RCS 

Leakage Instrumentation.’’ 

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 clarifies the 

operability requirements for the RCS leakage 

detection instrumentation and reduces the 

time allowed for the plant to operate when 

the only TS required operable RCS leakage 

detection instrumentation monitor is the 

containment atmosphere gaseous radiation 

monitor. The monitoring of RCS leakage is 

not a precursor to any accident previously 

evaluated. The monitoring of RCS leakage is 

not used to mitigate the consequences of any 

accident previously evaluated. The plant 

specific variation to this license amendment 

request, to insert the Note ‘‘Not required until 

12 hours after establishment of steady state 

operation’’ into applicable portions of the 

Technical Specification is administrative in 

nature. As a result, its inclusion does not 

impact any plant equipment’s ability to 

perform its required functions. Therefore, it 

is concluded that the proposed changes do 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change clarifies the 

operability requirements for the RCS leakage 

detection instrumentation and reduces the 

time allowed for the plant to operate when 

the only TS required operable RCS leakage 

detection instrumentation monitor is the 

containment atmosphere gaseous radiation 

monitor. The proposed change does not 

involve a physical alteration of the plant (no 

new or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. The 

proposed change maintains sufficient 

continuity and diversity of leak detection 

capability that the probability of piping 

evaluated and approved for leak-before-break 

progressing to pipe rupture remains 

extremely low. The plant specific variation to 

this license amendment request, to insert the 

Note ‘‘Not required until 12 hours after 

establishment of steady state operation’’ into 

applicable portions of the Technical 

Specification also does not involve a physical 

alteration of the plant or change in how plant 

equipment is operated. Therefore, it is 

concluded that the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

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Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 21921 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

The proposed change clarifies the 

operability requirements for the RCS leakage 

detection instrumentation and reduces the 

time allowed for the plant to operate when 

the only TS required operable RCS leakage 

detection instrumentation monitor is the 

containment atmosphere gaseous radiation 

monitor. Reducing the amount of time the 

plant is allowed to operate with only the 

containment atmosphere gaseous radiation 

monitor operable increases the margin of 

safety by increasing the likelihood that an 

increase in RCS leakage will be detected 

before it potentially results in gross failure. 

The plant specific variation to this license 

amendment request, to insert the Note ‘‘Not 

required until 12 hours after establishment of 

steady state operation’’ into applicable 

portions of the Technical Specification 

provides clarification as it reflects the time 

necessary for plant conditions to stabilize in 

order to ensure an accurate water inventory 

can be obtained. 

Therefore, it is concluded that 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 amendments 

request involves no significant hazards 

consideration. 

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 

amendments request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Generation Group, LLC, 

750 East Pratt Street, 17th floor, 

Baltimore, MD 21202. 

NRC Branch Chief: Nancy L. Salgado. 

Carolina Power and Light Company, 

Docket No. 50–261, H.B. Robinson 

Steam Electric Plant, Unit 2, Darlington 

County, South Carolina 

Date of amendment request: October 

20, 2010. 

Description of amendment request: 

The proposed amendment would revise 

the technical specifications (TS) 

description of fuel assemblies specified 

in TS 4.2.1. Additionally, changes are 

requested to the analytical methods 

referenced in TS 5.6.5.b. The changes to 

TS 5.6.5.b includes the addition of 

AREVA topical reports, BAW– 

10240(P)(A), ‘‘Incorporation of M5TM

Properties in Framatome ANP Approved 

Methods,’’ and EMF–2328(P)(A), ‘‘PWR 

Small Break LOCA Evaluation Model 

S–RELAP5 Based,’’ and the deletion of 

nine analytical methods that were 

previously approved but are no longer 

planned to be used, and therefore have 

not been analyzed for acceptability for 

M5TM (M5) alloy fuel. 

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 license amendment adds a 

Nuclear Regulatory Commission approved 

analytical method, BAW–10240(P)(A), 

‘‘Incorporation of M5TM Properties in 

Framatome ANP Approved Methods,’’ used 

to determine the core operating limits, to 

Technical Specification (TS) 5.6.5.b and 

changes the description of fuel assemblies 

specified in TS 4.2.1 to allow use of the M5 

alloy. The proposed amendment does not 

affect the acceptance criteria for any Final 

Safety Analysis Report (FSAR) safety analysis 

analyzed accidents or anticipated operational 

occurrences. The proposed amendment does 

not involve operation of the required 

structures, systems or components (SSCs) in 

a manner different from those previously 

recognized or evaluated. As such, the 

proposed amendment does not increase the 

probability or consequences of an accident. 

In addition, the proposed license 

amendment adds NRC approved 

methodology EMF–2328(P)(A), ‘‘PWR Small 

Break LOCA Evaluation Model, S–RELAP5 

Based.’’ This change, by itself, does not 

impact the current design bases. The 

proposed change enables the use of new 

methodologies to re-analyze small break lossof-coolant accidents. Revised analyses may 

either result in continued conformance 

within design bases, or may change the 

design bases. If design bases changes result 

from a revised analysis, then the specific 

design changes will be evaluated in 

accordance with HBRSEP, Unit 2, design 

change procedures and 10 CFR 50.59. 

Further, this part of the change does not 

involve physical changes to any plant 

structure, system, or component. 

In addition, the proposed license 

amendment deletes nine analytical methods 

that were previously approved and listed in 

Section 5.6.5.b, but are no longer planned to 

be used. This change is administrative in 

nature as it removes methodologies that have 

become obsolete and hence have not been 

analyzed for acceptability with M5 fuel. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would 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 Previously Evaluated. 

Use of M5 fuel will not result in changes 

in the operation or configuration of the 

facility. Topical reports BAW–10227(P)(A) 

and BAW–10240(P)(A) evaluate the material 

properties of the M5 alloy and conclude that 

they are similar or better than those of 

zircaloy-4. Therefore, M5 fuel rod cladding 

will perform similarly to those fabricated 

from zircaloy-4, thus precluding the 

possibility of the fuel becoming an accident 

initiator and causing a new or different type 

of accident. No new failure mechanisms will 

be introduced by the changes being 

requested. 

The proposed addition of EMF–2328(P)(A) 

does not involve any physical alteration of 

plant systems, structures, or components, 

other than allowing for fuel design in 

accordance with NRC-approved 

methodologies. No new or different 

equipment is being installed. No installed 

equipment is being operated in a different 

manner. There is no change to the parameters 

within which the plant is normally operated 

or in the setpoints that initiate protective or 

mitigative actions. As a result, no new failure 

modes are being introduced by introduction 

of this methodology. 

Therefore, the proposed amendment 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 change will not involve a 

significant reduction in the margin of safety 

because it has been demonstrated that the 

material properties of the M5 alloy are not 

significantly different from those of zircaloy4. M5 alloy is expected to perform similarly 

or better than zircaloy-4 for all normal 

operating and accident scenarios, including 

both loss-of-coolant accident (LOCA) and 

non-LOCA scenarios. The proposed changes 

do not affect the acceptance criteria for any 

FSAR safety analysis analyzed accidents or 

anticipated operational occurrences. All 

required safety limits would continue to be 

analyzed using methodologies approved by 

the Nuclear Regulatory Commission. 

There is no impact on any margin of safety 

resulting from the incorporation of these new 

topical reports into the Technical 

Specifications. If design basis changes result 

from a revised analysis that uses these new 

methodologies, the specific design changes 

will be evaluated in accordance with 

HBRSEP, Unit 2, design change procedures 

and 10 CFR 50.59. Any potential reduction 

in the margin of safety would be evaluated 

for that specific design change. 

Therefore, the proposed amendment would 

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 T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: Douglas A. 

Broaddus. 

mstockstill on DSKH9S0YB1PROD with NOTICES

21922 Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 

Carolina Power and Light Company, et 

al., Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of amendment request: January 

13, 2011. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specifications (TSs) to change 

the description of fuel assemblies 

specified in TS 5.3.1 and add the 

AREVA NP Inc., topical report, BAW– 

10240(P)(A), ‘‘Incorporation of M5TM

Properties in Framatome ANP Approved 

Methods,’’ to the referenced analytical 

methods in administrative TS 6.9.1.6.2 

to allow the use of M5TM alloy for fuel 

rod cladding. 

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 license amendment adds a 

NRC approved analytical method, BAW– 

10240(P)(A), ‘‘Incorporation of M5TM

Properties in Framatome ANP Approved 

Methods,’’ used to determine the core 

operating limits, to TS 6.9.1.6.2 and changes 

the description of fuel assemblies specified 

in TS 5.3.1 to allow use of the M5TM alloy. 

The proposed amendment does not affect the 

acceptance criteria for any Final Safety 

Analysis Report (FSAR) safety analysis 

analyzed accidents and anticipated 

operational occurrences. As such, the 

proposed amendment does not increase the 

probability or consequences of an accident. 

The proposed amendment does not involve 

operation of the required structures, systems 

or components (SSCs) in a manner or 

configuration different from those previously 

recognized or evaluated. Therefore, operation 

of the facility in accordance with the 

proposed amendment would 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. 

Use of M5TM clad fuel will not result in 

changes in the operation or configuration of 

the facility. Topical Report BAW–10240 

describes, by reference, that the material 

properties of the M5TM alloy are similar to or 

better than those of Zircaloy-4. Therefore, 

since M5TM fuel rod cladding will perform 

similarly to those fabricated from Zircaloy-4, 

the possibility of the fuel becoming an 

accident initiator and causing a new or 

different type of accident is precluded. Since 

the material properties of M5TM alloy are 

similar to or better than those of Zircaloy-4, 

there will be no significant changes in the 

types of any effluents that may be released 

off-site. There will not be a significant 

increase in occupational or public radiation 

exposure. The proposed amendment does not 

involve operation of any required SSCs in a 

manner or configuration different from those 

previously recognized or evaluated. No new 

failure mechanisms will be introduced by the 

changes being requested. Therefore, the 

proposed amendment does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change will not involve a 

significant reduction in the margin of safety 

because it has been demonstrated that the 

material properties of the M5TM alloy are not 

significantly different from those of Zircaloy4. M5TM alloy is expected to perform 

similarly to or better than Zircaloy-4 for all 

normal operating and accident scenarios, 

including both loss-of-coolant accident 

(LOCA) and non-LOCA scenarios. The 

proposed changes do not affect the 

acceptance criteria for any FSAR safety 

analysis analyzed accidents or anticipated 

operational occurrences. All required safety 

limits will continue to be analyzed using 

methodologies approved by the NRC. 

Therefore, the proposed amendment would 

not involve a significant reduction in a 

margin of safety. 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 T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: Douglas A. 

Broaddus. 

Carolina Power and Light Company, 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit 2, Darlington 

County, South Carolina 

Date of amendment request: January 

20, 2011. 

Description of amendment request: 

The proposed change would revise H. B. 

Robinson Steam Electric Plant 

Technical Specifications (TS) Section 

3.8.3, ‘‘Diesel Fuel Oil and Starting Air,’’ 

and Section 3.8.5, ‘‘DC Sources— 

Shutdown.’’ The proposed change to TS 

3.8.3 revises a nonconservative air 

receiver tank pressure to a value 

consistent with vendor 

recommendations. The proposed change 

to TS 3.8.5 corrects an editorial error 

related to TS formatting. 

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 to TS 3.8.3 revises a 

non-conservative value in the current TS for 

EDG air start pressure. The proposed value is 

consistent with vendor recommendations 

and will ensure that the intent of the TS 

requirement is met. Therefore, the proposed 

change will provide improved assurance that 

the EDGs will be able to meet their safety 

function. 

The proposed change to TS 3.8.5 is an 

editorial correction and there will be no 

actual changes to plant design or operation. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would 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 Previously Evaluated. 

As described above, the proposed change 

to TS 3.8.3 provides improved assurance that 

the EDGs will be able to meet their safety 

function. No new failure modes are 

introduced. Therefore, no new accident 

initiators or precursors are introduced by the 

proposed change. 

The proposed change to TS 3.8.5 is an 

editorial correction and there will be no 

actual changes to plant design or operation. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

3. The Proposed Change Does Not Involve 

a Significant Reduction in the Margin of 

Safety. 

As described above, the proposed change 

to TS 3.8.3 provides improved assurance that 

the EDGs will be able to meet their safety 

function of mitigating events that involve a 

loss of offsite power. Therefore, the proposed 

change will preserve any margin of safety. 

The proposed change to TS 3.8.5 is an 

editorial correction and there will be no 

actual changes to plant design or operation. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not involve a significant reduction in 

the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

mstockstill on DSKH9S0YB1PROD with NOTICES

Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 21923 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: Douglas A. 

Broaddus. 

Northern States Power Company— 

Minnesota, Docket No. 50–263, 

Monticello Nuclear Generating Plant 

(MNGP), Wright County, Minnesota 

Date of amendment request: February 

7, 2011. 

Description of amendment request: 

The licensee proposed to amend the 

MNGP Technical Specifications (TS), 

revising Surveillance Requirement 

3.5.1.7 regarding the Emergency Core 

Cooling System (ECCS) core spray flow 

from a minimum of 2800 gpm to a 

minimum of 2835 gpm. The licensee 

considers the current minimum flow 

rate requirement as non-conservative. 

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) analysis. The 

NRC staff reviewed the licensee’s NSHC 

analysis and has 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 purpose of the minimum core spray 

flow rate requirement is to ensure that the 

ECCS will perform as designed. Of the 

postulated accidents and transients 

previously analyzed in the MNGP Updated 

Safety Analysis Report, none of them were 

postulated to be initiated by the ECCS 

performing as designed. 

Furthermore, the consequences of the 

previously analyzed accidents were not 

postulated to be exacerbated by the ECCS 

performing as designed. Accordingly, the 

probability of occurrence and the 

consequences of the previously analyzed 

accidents would not be affected in any way 

by the proposed amendment to the TS. 

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 does not involve 

any physical alteration of the plant (no new 

or different type of equipment will be 

installed) nor does it change methods and 

procedures governing plant operation. The 

proposed amendment will not impose any 

new or eliminate any old requirements. 

Therefore, the proposed amendment does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment will not have 

any effect on previously used safety analysis 

methods, scenarios, acceptance criteria, or 

assumptions. 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: Robert J. 

Pascarelli. 

Virginia Electric and Power Company, 

Docket No. 50–281, Surry Power Station, 

Unit 2, Surry County, Virginia 

Date of amendment request: 

December 16, 2010. 

Description of amendment request: 

This amendment request proposes to 

revise Technical Specification (TS) 

6.4.Q, ‘‘Steam Generator (SG) Program,’’ 

to exclude portions of the SG tube 

below the top of the SG tubesheet from 

periodic tube inspections for Unit 2 

during Refueling Outage 23 and the 

subsequent operating cycle. This 

amendment request also proposes to 

revise TS 6.6.A.3, ‘‘Steam Generator 

Tube Inspection Report,’’ to provide 

reporting requirements specific to Unit 

2 for the temporary alternate repair 

criteria. 

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 previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components. The proposed 

change that alters the steam generator 

inspection/repair criteria and the steam 

generator inspection reporting criteria does 

not have a detrimental impact on the 

integrity of any plant structure, system, or 

component that initiates an analyzed event. 

The proposed change will not alter the 

operation of, or otherwise increase the failure 

probability of any plant equipment that 

initiates an analyzed accident. 

Of the applicable accidents previously 

evaluated, the limiting transients with 

consideration to the proposed change to the 

steam generator tube inspection and repair 

criteria are the steam generator tube rupture 

(SGTR) event and the steam line break (SLB) 

postulated accidents. 

During the SGTR event, the required 

structural integrity margins of the steam 

generator tubes and the tube-to-tubesheet 

joint over the H* distance will be 

maintained. Tube rupture in tubes with 

cracks within the tubesheet is precluded by 

the constraint provided by the tube-totubesheet joint. This constraint results from 

the hydraulic expansion process, thermal 

expansion mismatch between the tube and 

tubesheet, and from the differential pressure 

between the primary and secondary side. 

Based on this design, the structural margins 

against burst, as discussed in Regulatory 

Guide (RG) 1.121, ‘‘Bases for Plugging 

Degraded PWR [Pressurized-Water Reactor] 

Steam Generator Tubes,’’ are maintained for 

both normal and postulated accident 

conditions. 

The proposed change has no impact on the 

structural or leakage integrity of the portion 

of the tube outside of the tubesheet. The 

proposed change maintains structural 

integrity of the steam generator tubes and 

does not affect other systems, structures, 

components, or operational features. 

Therefore, the proposed change results in no 

significant increase in the probability of the 

occurrence of a SGTR accident. 

At normal operating pressures, leakage 

from primary water stress corrosion cracking 

below the proposed limited inspection depth 

is limited by both the tube-to-tubesheet 

crevice and the limited crack opening 

permitted by the tubesheet constraint. 

Consequently, negligible normal operating 

leakage is expected from cracks within the 

tubesheet region. The consequences of an 

SGTR event are affected by the primary to 

secondary leakage flow during the event. 

However, primary to secondary leakage flow 

through a postulated broken tube is not 

affected by the proposed changes since the 

tubesheet enhances the tube integrity in the 

region of the hydraulic expansion by 

precluding tube deformation beyond its 

initial hydraulically expanded outside 

diameter. Therefore, the proposed changes do 

not result in a significant increase in the 

consequences of a SGTR. 

The consequences of a steam line break 

(SLB) are also not significantly affected by 

the proposed changes. During a SLB 

accident, the reduction in pressure above the 

tubesheet on the shell side of the steam 

generator creates an axially uniformly 

distributed load on the tubesheet due to the 

reactor coolant system pressure on the 

underside of the tubesheet. The resulting 

bending action constrains the tubes in the 

tubesheet thereby restricting primary to 

secondary leakage below the midplane. 

Primary to secondary leakage from tube 

degradation in the tubesheet area during the 

limiting accident (i.e., a SLB) is limited by 

flow restrictions. These restrictions result 

from the crack and tube-to-tubesheet contact 

pressures that provide a restricted leakage 

path above the indications and also limit the 

degree of potential crack face opening as 

compared to free span indications. 

The probability of a SLB is unaffected by 

the potential failure of a steam generator tube 

as the failure of the tube is not an initiator 

for a SLB event. 

The leakage factor of 2.03 is a bounding 

value for all SGs, both hot and cold legs, in 

mstockstill on DSKH9S0YB1PROD with NOTICES

21924 Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 

Table 9–7 of WCAP–17092–P. Also as shown 

in Table 9–7 of WCAP–17092–P, for Surry for 

a postulated SLB, a leakage factor of 1.80 has 

been calculated. However, for Surry, a more 

conservative leakage factor of 2.03 will be 

applied to the normal operating leakage 

associated with the tubesheet expansion 

region in the condition monitoring (CM) 

assessment and the operational assessment 

(OA). Specifically, for the CM assessment, 

the component of leakage from the prior 

cycle from below the H* distance will be 

multiplied by a factor of 2.03 and added to 

the total leakage from any other source and 

compared to the allowable accident induced 

leakage limit. For the OA, the difference in 

the leakage between the allowable leakage 

and the accident induced leakage from 

sources other than the tubesheet expansion 

region will be divided by 2.03 and compared 

to the observed operational leakage. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change that alters the steam 

generator inspection/repair criteria and the 

steam generator inspection reporting criteria 

does not introduce any new equipment, 

create new failure modes for existing 

equipment, or create any new limiting single 

failures. Plant operation will not be altered, 

and all safety functions will continue to 

perform as previously assumed in accident 

analyses. 

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 change involve a significant 

reduction in a margin of safety? 

Response: No. 

The proposed change that alters the steam 

generator inspection/repair criteria and the 

steam generator inspection reporting criteria 

maintains the required structural margins of 

the steam generator tubes for both normal 

and accident conditions. NEI [Nuclear Energy 

Institute] 97–06, Revision 2, ‘‘Steam 

Generator Program Guidelines,’’ and RG 1.121 

are used as the bases in the development of 

the limited tubesheet inspection depth 

methodology for determining that steam 

generator tube integrity considerations are 

maintained within acceptable limits. RG 

1.121 describes a method acceptable to the 

NRC for meeting GDC [General Design 

Criteria] 14, ‘‘Reactor Coolant Pressure 

Boundary,’’ GDC 15, ‘‘Reactor Coolant System 

Design,’’ GDC 31, ‘‘Fracture Prevention of 

Reactor Coolant Pressure Boundary,’’ and 

GDC 32, ‘‘Inspection of Reactor Coolant 

Pressure Boundary,’’ by reducing the 

probability and consequences of a SGTR. RG 

1.121 concludes that by determining the 

limiting safe conditions for tube wall 

degradation the probability and 

consequences of a SGTR are reduced. This 

RG uses safety factors on loads for tube burst 

that are consistent with the requirements of 

Section III of the American Society of 

Mechanical Engineers (ASME) Code. 

For axially oriented cracking located 

within the tubesheet, tube burst is precluded 

due to the presence of the tubesheet. For 

circumferentially oriented cracking, the H* 

analysis, documented in Section 4 of the 

license amendment request, defines a length 

of degradation free expanded tubing that 

provides the necessary resistance to tube 

pullout due to the pressure induced forces, 

with applicable safety factors applied. 

Application of the limited hot and cold leg 

tubesheet inspection criteria will preclude 

unacceptable primary to secondary leakage 

during all plant conditions. The methodology 

for determining leakage provides for large 

margins between calculated and actual 

leakage values in the proposed limited 

tubesheet inspection depth criteria. 

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: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., 120 Tredegar 

St., RS–2, Richmond, VA 23219. 

NRC Branch Chief: Gloria Kulesa. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–449, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: May 18, 

2010, as supplemented by letter dated 

March 1, 2011. 

Brief description of amendment 

request: The proposed amendment 

would revise Technical Specification 

(TS) 6.8.3.I, ‘‘Containment PostTensioning System Surveillance 

Program.’’ TS 6.8.3.I states that the 

containment post-tensioning system 

surveillance program shall be in 

accordance with American Society of 

Mechanical Engineers (ASME) Code, 

Section XI, Subsection IWL, 1992 

Edition with 1992 Addenda, as 

supplemented by 10 CFR 

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

The proposed amendment removes 

the specific year of the applicable Code 

edition consistent with Revision 3.1 of 

NUREG–1431, ‘‘Standard Technical 

Specifications, Westinghouse Plants’’ 

and will allow for future updates to the 

surveillance program when the 

applicable code edition changes without 

requiring additional TS changes. 

Date of publication of individual 

notice in the Federal Register: March 

22, 2011 (76 FR 16012). 

Expiration date of individual notice: 

April 21, 2011 (public comments); May 

23, 2011 (hearing requests). 

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 

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Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 21925 

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, Room O1–F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland 

20852. Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

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

415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of application for amendments: 

March 31, 2010, as supplemented by 

letter dated November 30, 2010. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications to relocate specific 

surveillance frequencies to a licenseecontrolled program using a riskinformed justification. 

Date of issuance: March 29, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 263, 259. 

Renewed Facility Operating License 

Nos. NPF–35 and NPF–52: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in the Federal 

Register: November 16, 2010 (75 FR 

70034). The supplement dated 

November 30, 2010, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 29, 2011. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of application for amendments: 

March 24, 2010, as supplemented by 

letters dated November 18, 2010, and 

March 2, 2011. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications to relocate specific 

surveillance frequencies to a licenseecontrolled program using a riskinformed justification. 

Date of issuance: March 29, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 261, 241. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in the Federal 

Register: November 16, 2010 (75 FR 

70035). 

The supplements dated November 18, 

2010, and March 2, 2011, 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 March 29, 2011. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of application for amendment: 

March 15, 2010, as supplemented by 

letters dated August 30, 2010, 

September 21, 2010, January 31, 2011, 

and February 18, 2011. 

Brief description of amendment: This 

amendment request would modify the 

Technical Specifications to revise the 

setpoint and setpoint tolerances for 

safety relief valves (SRVs) and spring 

safety valves (SSVs) and support the 

plant modifications associated with the 

replacement of (1) four Target Rock twostage SRVs with three-stage SRVs, and 

(2) two existing Dresser 3.749 inch 

throat diameter SSVs with Dresser 4.956 

inch throat diameter SSVs. 

Date of issuance: March 28, 2011. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 235. 

Facility Operating License No. DPR– 

35: The amendment revised the License 

and Technical Specifications. 

Date of initial notice in the Federal 

Register: May 4, 2010 (75 FR 23812). 

The supplemental letters dated 

August 30, 2010, September 21, 2010, 

January 31, 2011, and February 18, 

2011, 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 this amendment is contained in a 

Safety Evaluation dated March 28, 2011. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of application for amendment: 

August 31, 2010. 

Brief description of amendments: The 

amendments involve administrative 

changes to the Technical Specifications 

(TSs). The changes involve: (1) Making 

an editorial change to Limerick 

Generating Station (LGS) Unit 1 TS 

Limiting Condition for Operation (LCO) 

3.3.1, Action b; (2) making an editorial 

change to LGS Units 1 and 2 TS Table 

3.3.1–1, Actions 2 and 9; (3) making the 

layout and format of LGS Unit 1 TS LCO 

3.6.5.3 Action requirements consistent 

with the LGS Unit 2 LCO Action 

requirements for the same TS; and 

(4) adding a reference to the minimum 

required number of operable main 

turbine bypass valves and the turbine 

bypass system response time to the core 

operating limits documented in the Core 

Operating Limits Report as specified in 

LGS, Units 1 and 2, TS 6.9.1.9. 

Date of issuance: March 31, 2011. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment Nos.: 200 and 161. 

Facility Operating License Nos. NPF– 

39 and NPF–85. The amendments 

revised the license and the technical 

specifications. 

Date of initial notice in the Federal 

Register: November 2, 2010 (75 FR 

67402). 

The Commission’s related evaluation 

of the amendment is contained in Safety 

Evaluation dated March 31, 2011. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–353, Limerick Generating 

Station, Unit 2, Montgomery County, 

Pennsylvania 

Date of application for amendment: 

December 15, 2010, as supplemented on 

February 17, 2011, and March 17, 2011. 

Brief description of amendment: The 

changes revise the Technical 

Specification (TS) relating to the Safety 

Limit Minimum Critical Power Ratios 

(SLMCPRs). The changes result from a 

cycle specific analysis performed to 

support the operation of Limerick 

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21926 Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 

Generating Station, Unit 2, in the 

upcoming Cycle 12. Specifically, the TS 

changes will revise the SLMCPRs 

contained in TS 2.1 for two 

recirculation loop operation and single 

recirculation loop operation to reflect 

the changes in the cycle specific 

analysis. The new SLMCPRs are 

calculated using Nuclear Regulatory 

Commission-approved methodology 

described in NEDE 24011–P–A, General 

Electric Standard Application for 

Reactor Fuel, Revision 17. 

Date of issuance: April 5, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 162. 

Facility Operating License No. NPF– 

85. The amendment revised the license 

and the technical specifications. 

Date of initial notice in the Federal 

Register: February 1, 2011 (76 FR 

5620). The supplements dated February 

17, 2011, and March 17, 2011, clarified 

the application, did not expand the 

scope of the application as originally 

noticed, and did not change the initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 5, 2011. 

No significant hazards consideration 

comments received: No. 

Indiana Michigan Power Company 

(IandM), Docket No. 50–316, Donald C. 

Cook Nuclear Plant, Unit 2, Berrien 

County, Michigan 

Date of application for amendment: 

March 19, 2009, as supplemented on 

November 20, 2009, February 24, March 

11, and March 25, 2011. 

Brief description of amendment: The 

amendment adopts a new analysis of a 

large-break loss-of-coolant accident, and 

revises the Technical Specifications to 

reflect this new analysis, which was 

performed using a plant-specific 

adaptation of the NRC-approved 

methodology set forth in Westinghouse 

Topical Report WCAP–16009–P–A, 

‘‘Realistic Large-Break LOCA Evaluation 

Methodology Using the Automated 

Statistical Treatment of Uncertainty 

Method (ASTRUM).’’ 

Date of issuance: March 31, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 297. 

Facility Operating License No. DPR– 

74: Amendment revised the Renewed 

Operating License and Technical 

Specifications. 

Date of initial notice in the Federal 

Register: August 11, 2009 (74 FR 

40238). 

The supplemental information 

contained clarifying information, did 

not change the scope of the license 

amendment request, did not change the 

NRC staff’s initial proposed finding of 

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 March 31, 2011. 

No significant hazards consideration 

comments received: No. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power Plant, 

Unit 1 and 2, Somervell County, Texas 

Date of amendment request: 

December 1, 2010. 

Brief description of amendments: The 

amendments revised the inspection 

scope and repair requirements in 

Technical Specification (TS) 5.5.9, ‘‘Unit 

1 Model D76 and Unit 2 Model D5 

Steam Generator (SG) Program,’’ to 

exclude portions of the Comanche Peak 

Nuclear Power Plant (CPNPP), Unit 2, 

Model D5 SG tubes below the top of the 

SG tubesheet from periodic SG tube 

inspections. In addition, the 

amendments revised TS 5.6.9, ‘‘Unit 1 

Model D76 and Unit 2 Model D5 Steam 

Generator Tube Inspection Reports,’’ to 

provide reporting requirements specific 

to CPNPP, Unit 2, for the temporary 

alternate repair criteria. The changes are 

applicable only to CPNPP, Unit 2, 

during Refueling Outage 12 and the 

subsequent operating cycle. 

Date of issuance: April 6, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to Mode 4 entry during startup 

from Unit 2 Refueling Outage 12. 

Amendment Nos.: Unit 1—154; Unit 

2—154. 

Facility Operating License Nos. NPF– 

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in the Federal 

Register: February 1, 2011 (76 FR 

5622). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 6, 2011. 

No significant hazards consideration 

comments received: No. 

NextEra Energy, Point Beach, LLC, 

Docket Nos. 50–266 and 50–301, Point 

Beach Nuclear Plant, Units 1 and 2, 

Town of Two Creeks, Manitowoc 

County, Wisconsin 

Date of application for amendments: 

April 7, 2009, as supplemented by 

letters dated June 17 (two letters), 

September 11, September 25, October 9, 

November 20 (two letters), November 21 

(two letters), November 30, December 8, 

and December 16 of 2009; and January 

7, January 8, January 22, February 11, 

February 25, March 3, April 15, April 

22, April 28, July 8, July 28, August 2, 

August 9, August 24, October 15, 

November 1, November 12 (two letters), 

November 30, and December 21 of 2010. 

The proposed changes were originally 

included as part of the April 7, 2009, 

extended power uprate (EPU) license 

amendment request, but subsequently 

divided into a separate licensing action 

for independent technical review. 

Brief description of amendments: The 

amendment changes the AFW system 

design and Technical Specifications 

(TS) 3.7.5, ‘‘Auxiliary Feedwater 

(AFW),’’ and TS 3.7.6, ‘‘Condensate 

Storage Tank (CST),’’ resulting from (1) 

modifications to the AFW system to 

support requirements for transients and 

other accidents at EPU conditions; (2) 

installation of main feedwater isolation 

valves to support accident mitigation by 

ensuring that containment pressure does 

not exceed safety analysis limits; (3) 

automatic AFW switchover from a CST 

suction source to a safety-related 

Service Water source; and (4) 

instrumentation setpoint changes 

supporting the aforementioned physical 

modifications. The upgrades and 

modifications to the AFW system are 

being installed to provide additional 

capacity and reliability for the system. 

Although the proposed changes are also 

designed to support the requirements 

for transients and other accidents at 

EPU conditions, the changes for this 

amendment have been evaluated using 

the current licensing basis. 

Date of issuance: March 25, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 180 days. 

Amendment Nos.: 238, 242. 

Renewed Facility Operating License 

Nos. DPR–24 and DPR–27: Amendments 

revise the License, Appendix C, and the 

Technical Specifications. 

Date of initial notice in the Federal 

Register: September 21, 2010 (75 FR 

57525). 

The supplemental letters contained 

clarifying information and did not 

change the staff’s initial proposed 

finding of no significant hazards 

consideration. 

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Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 21927 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 25, 2011. 

No significant hazards consideration 

comments received: No. 

NextEra Energy, Point Beach, LLC, 

Docket Nos. 50–266 and 50–301, Point 

Beach Nuclear Plant, Units 1 and 2, 

Town of Two Creeks, Manitowoc 

County, Wisconsin 

Date of application for amendments: 

April 7, 2009, as supplemented by 

letters dated June 17, September 11, 

September 25, November 20, November 

30, and December 8 of 2009; February 

11, February 25, April 22, April 30, July 

21, July 28, August 2, and September 28 

of 2010. 

Brief description of amendments: The 

amendment changed the Technical 

Specifications to support (1) 

modifications to the AFW system; (2) an 

EPU to increase plant core thermal 

power from 1,540 megawatts thermal 

(MWt) to 1,800 MWt; and (3) update 

non-conservative RPS and ESFAS 

setpoints not associated with the EPU. 

The amendment also modified the RPS 

instrumentation setpoints of TS Table 

3.3.1–1 and the ESFAS instrumentation 

setpoints of TS Table 3.3.2–1. The 

changes include both EPU and non-EPU 

related setpoints. The revised TS 

allowable values have been calculated 

to account for new analytical limits, 

instrument uncertainties, and 

instrument drift. The changes also 

include the addition of a new column 

entitled Nominal Trip Setpoint that was 

added to provide consistency with the 

TS Table format in NUREG 1431, 

‘‘Standard Technical Specifications— 

Westinghouse Plants,’’ and Technical 

Specification Task Force (TSTF)–493, 

Revision 4, ‘‘Clarify the Application of 

Setpoint Methodology for Limiting 

Safety System Setting (LSSS) 

Functions.’’ The RPS and ESFAS 

instrumentation uncertainty/setpoint 

calculations have also been revised to 

eliminate the use of a single-sided 

reduction factor in the total loop error 

determination for LSSS setpoints. 

Date of issuance: March 25, 2011. 

Effective date: As of the date of 

issuance, and shall be implemented 

prior to Unit 1 startup from the Fall 

2011 refueling outage (Unit 1) and 

within 180 days (Unit 2). 

Amendment Nos.: 239 and 243. 

Renewed Facility Operating License 

Nos. DPR–24 and DPR–27: Amendments 

revised the Technical Specifications/ 

License. 

Date of initial notice in the Federal 

Register: September 21, 2010 (75 FR 

57524). 

The supplemental letters contained 

clarifying information and did not 

change the staff’s initial proposed 

finding of no significant hazards 

consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 25, 2011. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of application for amendment: 

March 29, 2010, as supplemented by 

letters dated May 28, and September 30, 

2010, and two letters dated February 14, 

2011. 

Brief description of amendment: The 

amendment modifies the Technical 

Specifications (TSs) to extend the 

allowed outage time for the A and B 

emergency diesel generators from 72 

hours to 14 days. 

Date of issuance: March 25, 2011. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 188. 

Facility Operating License No. NPF– 

57: The amendment revised the TSs and 

the License. 

Date of initial notice in the Federal 

Register: June 29, 2010 (75 FR 37476). 

The letters dated May, 28, and 

September 30, 2010, and February 14, 

2011 (two letters), provided clarifying 

information that did not change the 

initial proposed no significant hazards 

consideration determination or expand 

the application beyond the scope of the 

original Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 25, 2011. 

No significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: April 13, 

2010, as supplemented by letters dated 

October 13 and December 21, 2010, and 

January 18, 2011. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) Table 3.3.2–1, 

‘‘Engineered Safety Feature Actuation 

System Instrumentation,’’ by adding a 

footnote to Function 8.a concerning the 

reactor trip P–4 engineered safety 

feature actuation system interlock. The 

footnote specifies which functions of 

the interlock are necessary in each mode 

in order to meet the limiting condition 

for operation. Specifically, the functions 

of tripping the main turbine and 

isolating main feedwater with a 

coincident low average temperature 

would no longer be applicable in MODE 

3, which is hot standby. Revised TS 

Table 3.3.2–1 also identifies that the 

function of the P–4 interlock that allows 

arming of the steam dump valves and 

transfers the steam dump load rejection 

(Tavg) controller to the plant trip 

controller is not required in any mode. 

Date of issuance: March 30, 2011. 

Effective date: The amendment will 

be effective upon issuance and will be 

implemented within 90 days from the 

date of issuance. 

Amendment No.: 194. 

Renewed Facility Operating License 

No. NPF–42. The amendment revised 

the Operating License and Technical 

Specifications. 

Date of initial notice in the Federal 

Register: June 15, 2010 (75 FR 33844). 

The supplemental letters dated October 

13 and December 21, 2010, and January 

18, 2011, 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 March 30, 2011. No 

significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

November 30, 2010. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 5.5.9, ‘‘Steam 

Generator (SG) Program,’’ to exclude 

portions of the tube below the top of the 

steam generator tubesheet from periodic 

steam generator tube inspections during 

Refueling Outage 18 and the subsequent 

operating cycle. In addition, TS 5.6.10, 

‘‘Steam Generator Tube Inspection 

Report’’ will be revised to remove a 

reference to the previous interim 

alternate repair criteria and to provide 

reporting requirements specific to the 

temporary alternate repair criteria. 

Date of issuance: April 6, 2011. 

Effective date: The amendment is 

effective upon issuance and will be 

implemented prior to MODE 4 entry 

during startup from Refueling Outage 

18. 

Amendment No.: 195. 

Renewed Facility Operating License 

No. NPF–42. The amendment revised 

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21928 Federal Register / Vol. 76, No. 75 / Tuesday, April 19, 2011 / Notices 

the Operating License and Technical 

Specifications. 

Date of initial notice in the Federal 

Register: February 1, 2011 (76 FR 

5623). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 6, 2011. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 8th day 

of April 2011. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2011–9177 Filed 4–18–11; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 50–027; NRC–2011–0083] 

Washington State University; Facility 

Operating License No. R–76; 

Washington State University Modified 

TRIGA Nuclear Radiation Center 

Reactor (NRCR); Environmental 

Assessment and Finding of No 

Significant Impact 

The U.S. Nuclear Regulatory 

Commission (NRC or the Commission) 

is considering the issuance of a renewed 

Facility Operating License No. R–76, to 

be held by Washington State University 

(WSU or the licensee), which would 

authorize continued operation of the 

Washington State University Modified 

TRIGA Nuclear Radiation Center 

Reactor (NRCR), located in the Dodgen 

Research Facility on Roundtop Drive in 

Pullman, Whitman County, Washington. 

Therefore, as required by Title 10 of the 

Code of Federal Regulations (10 CFR) 

§ 51.21, the NRC is issuing this 

Environmental Assessment and Finding 

of No Significant Impact. 

Environmental Assessment 

Identification of Proposed Action 

The proposed action would renew 

Facility Operating License No. R–76 for 

a period of twenty years from the date 

of issuance of the renewed license. The 

proposed action is in accordance with 

the licensee’s application dated June 24, 

2002, as supplemented by letters dated 

August 15, 2007, June 13, 2008, and 

April 7, 2010. In accordance with 10 

CFR 2.109, the existing license remains 

in effect until the NRC takes final action 

on the renewal application. 

Need for the Proposed Action 

The proposed action is needed to 

allow the continued operation of the 

NRCR to routinely provide teaching 

opportunities, services and research for 

numerous institutions for a period of 

twenty years. 

Environmental Impacts of the Proposed 

Action 

The NRC has completed its safety 

evaluation of the proposed action to 

issue a renewed Facility Operating 

License No. R–76 to allow continued 

operation of the NRCR for a period of 

twenty years and concludes there is 

reasonable assurance that the NRCR will 

continue to operate safely for the 

additional period of time. The details of 

the NRC staff’s safety evaluation will be 

provided with the renewed license that 

will be issued as part of the letter to the 

licensee approving its license renewal 

application. This document contains the 

environmental assessment of the 

proposed action. 

The NRCR is located 1.27 kilometers 

(0.79 miles) east of the French 

Administration Building on the main 

campus of WSU. The NRCR is located 

in the Dodgen Research Facility. The 

Dodgen Research Facility is a multipurpose building constructed primarily 

of concrete, brick, steel, and aluminum. 

The entrance to the Dodgen Research 

Facility is secured and an access code 

is required for entry. Emergency exit 

doors in the Dodgen Research Facility 

are key-locked from the outside and 

only a few individuals are issued the 

key. Entry into the NRCR from the 

Dodgen Research Facility requires a 

special key or confirmation of identity 

through closed-circuit television and 

verbal contact with the reactor 

operators. There are three outside 

entrances allowing direct access to the 

NRCR. These entrances are secured and 

the area around each one is surrounded 

by a fence and jersey barriers. The 

exclusion zone is considered to be the 

perimeter of the reactor building. A road 

and unused land is located west of the 

site. Until late 2008, the site was 

surrounded for a distance of 400 meters 

(1300 feet) in all directions by grazing 

land for livestock which was owned by 

WSU. The land has since been 

converted into a golf course which 

surrounds the NRCR in all directions 

except the west. The land remains 

uninhabited. The golf course is 

separated from the NRCR by 100 to 200 

meters (330 to 660 feet) of land. There 

is a parcel of land abutting the NRCR of 

about 10,000 square meters (109,000 

square feet) of virgin prairie land which, 

by regulation or policy, WSU has no 

plans to use. The closest building is 411 

meters (1350 feet) west of the NRCR. 

The closest occupied dwellings are 626 

meters (2060 feet) to the westsouthwest. 

The NRCR is a pool-type, light water 

moderated and cooled research reactor 

licensed to operate at a maximum 

steady-state power level of 1 megawatt 

thermal power (MW(t)). The reactor is 

also licensed to operate in a pulse mode 

to a peak power of approximately 2,000 

MW(t). The fuel is contained in a reactor 

vessel suspended from a movable bridge 

and is located near the bottom of an 8 

meter (25 feet) deep concrete pool 

containing approximately 242,000 liters 

(63,930 gallons) of water. The reactor is 

fueled with standard low-enriched 

uranium TRIGA (Training, Research, 

Isotopes, General Atomic) fuel. A 

detailed description of the reactor can 

be found in the NRCR Safety Analysis 

Report (SAR). There have been two 

major modifications to the Facility 

Operating License since renewal of the 

license on August 11, 1982. Orders were 

issued: (1) Allowing for an increase in 

the possession limits for Uranium-235; 

and (2) conversion from high-enriched 

uranium fuel to low-enriched uranium 

fuel as amendments to the license. 

The licensee has not requested any 

changes in the NRCR design or 

operating conditions as part of the 

application for license renewal. No 

changes are being made in the types or 

quantities of effluents that may be 

released off site. The licensee has 

systems in place for controlling the 

releases of radiological effluents and 

implements a radiation protection 

program to monitor personnel exposures 

and releases of radioactive effluents. 

Accordingly, there would be no increase 

in routine occupational or public 

radiation exposure as a result of the 

license renewal. As discussed in the 

NRC staff’s safety evaluation, the 

proposed action will not significantly 

increase the probability or consequences 

of accidents. Therefore, license renewal 

would not change the environmental 

impact of NRCR operation. The NRC 

staff evaluated information contained in 

the licensee’s application and data 

reported to the NRC by the licensee for 

the last five years of operation to 

determine the projected radiological 

impact of the NRCR on the environment 

during the period of the renewed 

license. The NRC staff finds that 

releases of radioactive material and 

personnel exposures were all well 

within applicable regulatory limits. 

Based on this evaluation, the NRC staff 

concludes that continued operation of 

the reactor would not have a significant 

environmental impact. 

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