Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 81667 

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

NTSB home page at http:// 

www.ntsb.gov. 

FOR MORE INFORMATION CONTACT: Candi 

Bing, (202) 314–6403. 

Dated: December 23, 2010. 

Candi R. Bing, 

Federal Register Liaison Officer. 

[FR Doc. 2010–32779 Filed 12–23–10; 4:15 pm] 

BILLING CODE 7533–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0393] 

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 December 2, 

2010, to December 15, 2010. The last 

biweekly notice was published on 

December 14, 2010 (75 FR 77906). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), 50.92, this means 

that operation of the facility in 

accordance with the proposed 

amendment would not (1) Involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules, 

Announcements and Directives Branch 

(RADB), TWB–05–B01M, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be faxed to the RADB at 301–492– 

3446. Documents may be examined, 

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

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR Part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area O1 F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 


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81668 Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

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

accordance with the NRC E-Filing rule 

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

submit and serve all adjudicatory 

documents over the Internet, or in some 

cases to mail copies on electronic 

storage media. Participants may not 

submit paper copies of their filings 

unless they seek an exemption in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

(10) days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by e-mail at 

hearing.docket@nrc.gov, or by telephone 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

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

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on 

NRC’s public Web site at http:// 

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

apply-certificates.html. System 

requirements for accessing the ESubmittal server are detailed in NRC’s 

‘‘Guidance for Electronic Submission,’’ 

which is available on the agency’s 

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

site-help/e-submittals.html. Participants 

may attempt to use other software not 

listed on the Web site, but should note 

that the NRC’s E-Filing system does not 

support unlisted software, and the NRC 

Meta System Help Desk will not be able 

to offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through EIE, users will be 

required to install a Web browser plugin from the NRC Web site. Further 

information on the Web-based 

submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC public Web site at 

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

complete at the time the documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

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

Time on the due date. Upon receipt of 

a transmission, the E-Filing system 

time-stamps the document and sends 

the submitter an e-mail notice 

confirming receipt of the document. The 

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

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the agency’s adjudicatory E-Filing 

system may seek assistance by 

contacting the NRC Meta System Help 

Desk through the ‘‘Contact Us’’ link 

located on the NRC Web site at http:// 

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

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

Meta System Help Desk is available 

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

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852, Attention: Rulemaking 

and Adjudications Staff. Participants 

filing a document in this manner are 

responsible for serving the document on 

all other participants. Filing is 

considered complete by first-class mail 

as of the time of deposit in the mail, or 

by courier, express mail, or expedited 

delivery service upon depositing the 

document with the provider of the 

service. A presiding officer, having 

granted an exemption request from 

using E-Filing, may require a participant 

or party to use E-Filing if the presiding 

officer subsequently determines that the 

reason for granting the exemption from 

use of E-Filing no longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

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

unless excluded pursuant to an order of 

the Commission, or the presiding 

officer. Participants are requested not to 

include personal privacy information, 

such as social security numbers, home 


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Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 81669 

addresses, or home phone numbers in 

their filings, unless an NRC regulation 

or other law requires submission of such 

information. With respect to 

copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

Petitions for leave to intervene must 

be filed no later than 60 days from the 

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

absent a determination by the presiding 

officer that the petition or request 

should be granted or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

For further details with respect to this 

license amendment application, see the 

application for amendment which is 

available for public inspection at the 

Commission’s PDR, located at One 

White Flint North, Public File Area O1– 

F21, 11555 Rockville Pike (first floor), 

Rockville, Maryland. Publicly available 

records will be accessible from the 

ADAMS Public Electronic Reading 

Room on the Internet at the NRC Web 

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

adams.html. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units 1, 2, and 3, 

Maricopa County, Arizona 

Date of amendment request: July 22, 

2010. 

Description of amendment request: 

The amendments would revise an 

element of the methodology used in 

evaluating the radiological 

consequences of design basis steam 

generator tube rupture (SGTR) 

accidents. Specifically, the changes will 

revise the Palo Verde Nuclear 

Generating Station (PVNGS) Updated 

Final Safety Analysis Report (UFSAR), 

Section 15.6.6, ‘‘Steam Generator Tube 

Rupture,’’ to reflect a lower iodine 

spiking factor assumed for the 

coincident event Generated Iodine 

Spike (GIS) and the resulting reduction 

in the radiological consequences 

provided in UFSAR Table 15.6.3–5, 

‘‘Radiological Consequences for the 

Limiting SGTRLOPSF [Steam Generator 

Tube Rupture with Loss of Offsite 

Power and Single Failure] Event.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment changes an 

element of the methodology used in 

evaluating the radiological consequences of 

design basis SGTR accidents. This change 

will revise the iodine spiking factor used for 

a GIS from a value of 500 to a value of 335. 

The proposed change in the methodology 

element does not involve any design or 

physical changes to the facility or any 

component of that facility. The proposed 

change creates no new failure modes or 

initiating occurrences that could result in a 

design basis transient or accident evaluated 

in the Palo Verde Nuclear Generating Station 

(PVNGS) Updated Final Safety Analysis 

Report (UFSAR). Therefore the proposed 

change does not involve a significant 

increase in the probability of an accident 

previously evaluated. 

The proposed change in the methodology 

element does change the design basis 

analyses results for PVNGS. However, the 

results remain bounded by the previous 

analyzed values and remain within the 

acceptance criteria for PVNGS of 100% of the 

10 CFR [Part] 100 maximum thyroid dose 

limit of 300 rem [roentgen equivalent man]. 

Therefore, the proposed change does not 

involve a significant increase in the 

consequences of an accident previously 

analyzed. 

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 changes an 

element of the methodology used in 

evaluating the radiological consequences of 

design basis SGTR accidents. This change 

will revise the iodine spiking factor used for 

a GIS from a value of 500 to a value of 335. 

The proposed change in the methodology 

element does not involve any design or 

physical changes to the facility or any 

component of that facility. The proposed 

change in the methodology element does 

change the design basis analyses results for 

PVNGS; however, these results remain 

bounded by the previous analyzed values 

and remain within the acceptance criteria for 

PVNGS of 100% of the 10 CFR [Part] 100 

maximum thyroid dose limit of 300 rem. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment changes an 

element of the methodology used in 

evaluating the radiological consequences of 

design basis SGTR accidents. This change 

will revise the iodine spiking factor used for 

a GIS from a value of 500 to a value of 335. 

The proposed change in the methodology 

element does not involve any design or 

physical changes to the facility or any 

component of that facility. The proposed 

methodology element change for a postulated 

SGTR, with a coincident loss of offsite 

power, GIS, and a failed open atmospheric 

dump valve (ADV), results in lower 

maximum dose consequences at the 

Exclusion Area Boundary (EAB) and Low 

Population Zone (LPZ) [than] previously 

analyzed for this event combination. The 

methodology element change results in the 2- 

hour maximum thyroid dose value of 182 

rem at the EAB being reduced to 124 rem. In 

addition, the 8-hour maximum thyroid dose 

of 125 rem at the LPZ, would be reduced to 

84 rem. 

Previously for PVNGS, the GIS 8-hour 

maximum thyroid dose was bounding at the 

LPZ and the pre-Accident Iodine Spike (PIS) 

2-hour maximum thyroid dose was bounding 

at the EAB. The methodology element change 

reduces the GIS calculated dose at both the 

EAB and LPZ for SGTR events, but it does 

not affect the PIS dose values. Since the GIS 

calculated dose at the LPZ drops below the 

PIS 8-hour LPZ maximum thyroid dose (91 

rem), the PIS 8-hour LPZ dose will become 

bounding for PVNGS. The PIS 2-hour EAB 

maximum thyroid dose (294 rem), remains 

the bounding dose at the EAB. 

The revised dose consequences remain 

bounded by the previous analyzed values 

and remain within the 10 CFR Part 100 

guideline values which are the acceptance 

criteria for PVNGS Units 1, 2, and 3. In 

addition, the proposed change has no effect 

on previously reported dose consequences 

for control room personnel following any 

postulated SGTR event. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on that 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the request 

for amendments involves no significant 

hazards consideration. 

Attorney for licensee: Michael G. 

Green, Senior Regulatory Counsel, 

Pinnacle West Capital Corporation, P.O. 

Box 52034, Mail Station 8695, Phoenix, 

Arizona 85072–2034. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit No. 3, 

Westchester County, New York 

Date of amendment request: October 

6, 2010. 

Description of amendment request: 

The proposed change will revise the 

note in Surveillance Requirement (SR) 

3.5.4.1 in the Refueling Water Storage 


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81670 Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 

Tank (RWST) Technical Specification 

(TS). Specifically, the proposed change 

will not require monitoring of the RWST 

temperature every 24 hours when the 

RWST heating steam supply isolation 

valves are locked closed. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change revises the 

existing Indian Point [Nuclear Generating 

Unit No.] 3 [(IP3)] Refueling Water Storage 

Tank (RWST) Technical Specification (TS) 

Surveillance Requirement (SR) 3.5.4.1 to 

revise the note that eliminates the 

requirement to perform SR 3.5.4.1 when 

ambient air temperatures are within the 

operating limits of the RWST. The revision 

to the note adds a requirement that the steam 

heating supply isolation valves be locked 

closed when not performing the surveillance. 

The additional requirement does not increase 

the probability of an accident occurring since 

it is not an accident initiator and does not 

increase the consequences of an accident 

since it is providing additional assurance that 

the RWST is within the temperature limits 

assumed for accident analyses. The change 

increases observation of the RWST 

temperature when the steam supply isolation 

valves are not locked closed and does not 

otherwise affect [* * *] the performance 

capability of the structures, systems, and 

components relied upon to mitigate the 

consequences of postulated accidents. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed change revises the note 

that eliminates the requirement to perform 

SR 3.5.4.1 when ambient air temperatures are 

within the operating limits of the RWST. The 

revision adds the additional requirement of 

locking closed the steam supply isolation 

valves. The proposed change does not 

involve installation of new equipment or 

modification of existing equipment, so that 

no new equipment failure modes are 

introduced. Also, the proposed change does 

not result in a change to the way that the 

equipment or facility is operated so that no 

new accident initiators are created. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The proposed change revises the note 

that eliminates the requirement to perform 

SR 3.5.4.1 when ambient air temperatures are 

within the operating limits of the RWST. The 

revision adds the additional requirement of 

locking closed the steam supply isolation 

valves. The change does not reduce margin 

since it increases the temperature 

surveillance frequency for the RWST to 

provide further assurance that the required 

water temperature is maintained at all times. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Operations, Inc., System 

Energy Resources, Inc., South 

Mississippi Electric Power Association, 

and Entergy Mississippi, Inc., Docket 

No. 50–416, Grand Gulf Nuclear 

Station, Unit 1, Claiborne County, 

Mississippi 

Date of amendment request: 

November 8, 2010. 

Description of amendment request: 

The proposed license amendment 

request will make changes related to the 

final resolution of an unresolved issue 

associated with Technical Specification 

(TS) Amendment No. 181 dated 

February 25, 2009. This issue was 

resolved with the approval of Revision 

4 of Technical Specification Task Force 

(TSTF) Change Traveler TSTF–493, 

‘‘Clarify Application of Setpoint 

Methodology for LSSS [Limiting Safety 

System Setting] Functions,’’ which 

included the instrument function (i.e., 

Condensate Storage Tank (CST) LevelLow) that was the subject of 

Amendment No. 181. Specifically, the 

proposed change will add the 

appropriate notes as specified in TSTF– 

493 to the surveillance requirements 

associated with TS Table 3.3.5.1–1, 

‘‘Emergency Core Cooling System 

Instrumentation,’’ Function 3.d, 

Condensate Storage Tank Level—Low, 

and to TS Table 3.3.5.2–1, ‘‘Reactor Core 

Isolation Cooling System 

Instrumentation,’’ Function 3, 

Condensate Storage Tank Level—Low. 

The supporting TS Bases will also be 

revised. 

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 adds test 

requirements to the CST Level-Low function 

to ensure the CST Level-low instruments will 

function as required. Surveillance tests are 

not an initiator of any accident previously 

evaluated. The CST components, for which 

the additional requirements were added, 

continue to be operable and capable of 

performing their intended function. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

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

different type of equipment will be installed. 

The proposed change does not alter 

assumptions made in the safety analysis but 

ensures that the CST Level-low instruments 

perform as assumed in the [Updated Final 

Safety Analysis Report]. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change adds test 

requirements that will assure that (1) the CST 

Level-low instrumentation for the setpoint 

allowable value will be the limiting setting 

for assessing instrumentation channel 

operability and (2) will be conservatively 

determined so that the evaluation of CST 

instrument performance history and the 

requirements of the calibration procedures 

will not have an adverse effect on equipment 

operability. The testing methods and 

acceptance criteria for the CST Level-low 

instrumentation will continue to be met. 

There is no impact to the safety analysis 

acceptance criteria as described in the plant 

licensing basis because no change is made to 

the accident analysis assumptions. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 


emcdonald on DSK2BSOYB1PROD with NOTICES
Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 81671 

NRC Branch Chief: Michael T. 

Markley. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: July 12, 

2010. 

Description of amendment request: 

The proposed amendment would 

modify Item 1 of Table 2–5, 

‘‘Instrumentation Operating 

Requirements for Other Safety Feature 

Functions,’’ of Technical Specification 

(TS) 2.15, ‘‘Instrumentation and Control 

Systems,’’ to provide new Note (e), and 

Surveillance Requirement (SR) Items 1 

and 2 of Table 3–3, ‘‘Minimum 

Frequencies for Checks, Calibrations 

and Testing of Miscellaneous 

Instrumentation and Controls,’’ of TS 

3.1, ‘‘Instrumentation and Control,’’ 

which pertain to operability of the 

primary and secondary control element 

assembly (CEA) position indication 

system (CEAPIS) channels. A new SR is 

proposed for Item 4 of Table 3–3 of TS 

3.1, which will verify the position of 

CEAs each shift. The proposed 

amendment will ensure that CEA 

alignment is maintained during power 

operations so that the power 

distribution and reactivity limits 

defined by the design power peaking 

and shutdown margin (SDM) limits are 

preserved. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment will allow plant 

operation to continue when a CEAPIS 

channel is inoperable by requiring prompt 

verification of CEA positions following CEA 

movement. CEAs are most likely to become 

misaligned during movement and therefore, 

this change will cause CEA alignment errors 

to be promptly detected and corrected. It is 

appropriate to clarify that CEAPIS channels 

are not subject to the requirements of TS 

2.15(1), (2), and (3) as they are not designed 

to be placed in trip or bypass, nor are they 

engineered safety feature (ESF) or isolation 

logic subsystems. 

The proposed amendment does not alter 

the requirements of TS 2.15(4) regarding the 

rod block function of the secondary CEAPIS 

channel. Should the secondary CEAPIS 

channel or its rod block function be 

inoperable, several additional CEA deviation 

events are possible. However, this situation 

is already addressed by TS 2.15(4), which 

requires the CEAs (rods) to be maintained 

fully withdrawn with the control rod drive 

system mode switch in the off position 

except when manual motion of CEA Group 

4 is required to control axial power 

distribution. This is the same position that 

the CEAs must be in (fully withdrawn) when 

the plant is at power (Mode 1) in order to 

utilize distributed control system (DCS) core 

mimic to CHANNEL CHECK the CEAPIS 

channels. 

If it was not possible to use DCS core 

mimic to verify the primary CEAPIS channel 

as would be the case if CEA Group 4 was 

inserted to control axial power distribution, 

then the primary CEAPIS channel would be 

declared inoperable when the CHANNEL 

CHECK could not be accomplished. The 

plant would then be placed in hot shutdown 

(Mode 3) within 12 hours in accordance with 

TS 2.15(4). Therefore, although the proposed 

amendment will allow a CEAPIS channel to 

be inoperable indefinitely, there is no 

significant increase in the probability or 

consequences of an accident as the 

requirements of TS 2.15(4) will continue to 

be met. This serves to prevent the type of 

CEA deviation events that the rod block 

function was designed for. 

Replacing the current method of verifying 

CEAPIS data with the defined term 

CHANNEL CHECK is an improvement that 

provides additional flexibility without 

weakening the intent of the surveillance. As 

a result, when it is feasible to obtain CEA 

position indication from DCS core mimic 

(i.e., when the CEAs are either fully inserted 

or fully withdrawn), the primary and 

secondary CEAPIS channels will be 

compared with DCS core mimic indication as 

well as each other. 

As an additional means of verifying CEA 

positions, DCS core mimic indication 

provides added confidence that the CEAs are 

in the indicated positions. Should the 

primary or secondary CEAPIS channel 

become inoperable, the accuracy and 

reliability of DCS core mimic indication is 

assured by its previous comparison with both 

OPERABLE channels. Comparison of the 

OPERABLE CEAPIS channel with DCS core 

mimic will satisfy the required CHANNEL 

CHECK and allow continued operation while 

the inoperable channel is repaired. The 

proposed amendment ensures that the CEA 

alignment required by TS 2.10.2(4) is met 

each shift by requiring all full length 

(shutdown and regulating) CEAs to be 

positioned within 12 inches of all other CEAs 

in the group. 

The change proposed for TS 2.10.2(7)c 

incorporates more conservative wording to 

ensure that the regulating CEA groups are 

maintained within the Long Term Insertion 

Limit. The proposed change will ensure that 

corrective actions are taken if either time 

interval is exceeded and makes TS 2.10.2(7)c 

more consistent with CE STS. 

The proposed amendment does not alter 

the plant configuration, require new plant 

equipment to be installed, alter accident 

analysis assumptions, add any initiators, or 

affect the function of plant systems or the 

manner in which systems are operated, 

maintained, modified, tested, or inspected. 

As an additional means of verifying primary 

and secondary CEAPIS data, DCS core mimic 

indication increases confidence in the 

reliability of CEAPIS data. 

The proposed amendment will help 

minimize unplanned shutdowns that can 

cause plant transients yet continues to ensure 

that power distribution and reactivity limits 

are maintained. 

Therefore, it is concluded that this change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment does not change 

the design function or operation of the 

primary or secondary CEAPIS channels. If 

one CEAPIS channel should become 

inoperable, the position of CEAs will be 

verified within 15 minutes of any CEA 

movement to quickly detect and correct CEA 

alignment errors. Data from each CEAPIS 

channel will continue to be compared to the 

other channel each shift as before. However, 

a CHANNEL CHECK will require that 

CEAPIS channel data also be compared with 

DCS core mimic indication when it is 

available. Thus, when the CEAPIS channels 

are required to be OPERABLE, there will be 

at least two means of verifying the position 

of CEAs or else appropriate actions must be 

taken. The CEA alignment required by TS 

2.10.2(4) is assured by requiring verification 

each shift that all full length (shutdown and 

regulating) CEAs are positioned within 12 

inches of all other CEAs in the group. 

No changes are proposed to testing and 

calibration of the CEAPIS channels and these 

requirements will continue to ensure that 

they are capable of performing their design 

function. Use of the defined term CHANNEL 

CHECK is an appropriate surveillance 

method as it requires that the channel be 

compared with other independent channels 

measuring the same variable where feasible. 

DCS core mimic is a diverse, accurate and 

reliable means of verifying CEA positions 

when the CEAs are fully inserted or fully 

withdrawn. The change proposed for TS 

2.10.2(7)c ensures that appropriate corrective 

actions are taken when the regulating CEA 

groups are below the Long Term Insertion 

Limit in excess of either of the specified time 

intervals. 

No new structures, systems, or components 

(SSCs) are being installed, and no credible 

new failure mechanisms, malfunctions, or 

accident initiators are created. 

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. 

When a CEAPIS channel is inoperable, the 

proposed amendment allows plant operation 

to continue but requires more frequent 

verification of CEA positions following any 

CEA movement, which is when CEAs are 

most likely to become misaligned. This will 

enable CEA alignment errors to be detected 

and corrected more promptly. As CEAPIS 

channels are not designed to be placed in trip 


emcdonald on DSK2BSOYB1PROD with NOTICES
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or bypass, nor are they engineered safety 

feature (ESF) or isolation logic subsystems, it 

is appropriate to clarify that TS 2.15(1), (2), 

and (3) do not apply. FCS normally operates 

with the CEAs fully withdrawn and 

maintains reactivity control by adjusting 

reactor coolant system (RCS) boric acid 

concentration. When the CEAs are fully 

withdrawn (or fully inserted), DCS core 

mimic indication provides accurate and 

reliable indication of CEA positions suitable 

for comparison with the primary and 

secondary CEAPIS channels. Thus, even with 

one CEAPIS channel inoperable, a diverse 

means of verifying the accuracy of the 

OPERABLE CEAPIS channel will be 

available. The accuracy and reliability of DCS 

core mimic is assured by testing conducted 

each refueling outage with continued 

assurance provided by comparison with 

primary and secondary CEAPIS each shift. 

The change also ensures that the CEA 

alignment required by TS 2.10.2(4) is met 

each shift by requiring all full length 

(shutdown and regulating) CEAs to be 

positioned within 12 inches of all other CEAs 

in the group. The proposed amendment does 

not alter the TS 2.15(4) requirement to place 

the reactor in hot shutdown in the event that 

both CEAPIS channels are inoperable. The 

change proposed for TS 2.10.2(7)c 

incorporates more conservative wording to 

ensure that the regulating CEA groups are 

maintained within the Long Term Insertion 

Limit. 

The proposed amendment will help 

minimize unplanned shutdowns that can 

cause plant transients yet continues to ensure 

that power distribution and reactivity limits 

are maintained. The proposed amendment 

does not alter the plant configuration, require 

new plant equipment to be installed, alter 

accident analysis assumptions, add any 

initiators, or affect the function of plant 

systems or the manner in which systems are 

operated, maintained, modified, tested, or 

inspected. 

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 this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David A. Repka, 

Esq., Winston & Strawn, 1700 K Street, 

NW., Washington, DC 20006–3817. 

NRC Branch Chief: Michael T. 

Markley. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of amendment request: October 

4, 2010. 

Description of amendment request: 

The proposed amendment would 

modify the Technical Specification (TS) 

requirements for snubbers in TS 3/4.7.9 

due to planned revisions to the 

inservice inspection (ISI) program. 

For the current third 10-year ISI 

intervals, at Salem Nuclear Generating 

Station (Salem), Units 1 and 2, snubber 

testing and examination are performed 

in accordance with the specific 

requirements of TS 3/4.7.9 in lieu of the 

requirements contained in American 

Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel Code 

(Code), Section XI, Article IWF–5000, as 

previously authorized by the U.S. 

Nuclear Regulatory Commission (NRC 

or the Commission). 

Section 50.55a(g)(4)(ii) of Title 10 of 

the Code of Federal Regulations (10 

CFR) requires that inservice 

examination of components conducted 

during successive 120-month inspection 

intervals must comply with the 

requirements of the latest edition and 

addenda of the ASME Code 

incorporated by reference in 10 CFR 

50.55a(b), 12 months before the start of 

the inspection interval. For the Salem 

Unit 1 fourth 10-year ISI interval 

beginning on May 20, 2011, the licensee 

intends to adopt Subsection ISTD of the 

ASME Code for Operation and 

Maintenance of Nuclear Power Plants 

(OM Code), 2004 Edition, in place of the 

requirements for snubbers in ASME 

Code, Section XI, Articles IWF–5200(a) 

and (b) and IWF–5300(a) and (b), as 

permitted by 10 CFR 50.55a(b)(3)(v). 

The licensee also intends to adopt 

Subsection ISTD of the ASME OM Code 

for the remainder of the Salem Unit 2 

third 10-year ISI interval which ends on 

November 27, 2013. 

In accordance with 10 CFR 

50.55a(g)(5)(ii), if a revised ISI program 

for a facility conflicts with the TSs for 

the facility, licensees are required to 

apply to the Commission for 

amendment of the TSs to conform the 

TSs to the revised program. Due to the 

planned changes to the ISI program, the 

proposed amendment would replace the 

specific TS requirements for snubbers, 

currently contained in surveillance 

requirement (SR) 4.7.9, with reference to 

the program for examination, testing 

and service life monitoring for snubbers. 

In addition, the current reference to SR 

4.7.9c in TS ACTION 3.7.9 would be 

replaced with reference to the program 

for snubbers. 

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, with Nuclear Regulatory 

Commission (NRC) staff edits in square 

brackets: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise [TS 3/4.7.9 

due to planned changes to the ISI program for 

snubbers. Specifically, the proposed 

amendment would replace the TS SRs for 

snubbers with reference to the program for 

examination, testing and service life 

monitoring for snubbers. Following 

implementation of the proposed amendment, 

in lieu of the TS SRs, snubber examination, 

testing and service life monitoring would be 

governed by the requirements in Section XI 

of the ASME Code or the OM Code as 

required by 10 CFR 50.55a(g) or 10 CFR 

.55a(b)(3)(v), except where the NRC has 

granted specific written relief, pursuant to 10 

CFR 50.55a(g)(6)(i), or authorized alternatives 

pursuant to 10 CFR 50.55a(a)(3).] 

Snubber examination, testing and service 

life monitoring is not an initiator of any 

accident previously evaluated. Therefore, the 

probability of an accident previously 

evaluated is not significantly increased. 

Snubbers will continue to be demonstrated 

OPERABLE by performance of a program for 

examination, testing and service life 

monitoring in compliance with 10 CFR 

50.55a or authorized alternatives. The 

proposed change to TS ACTION 3.7.9 for 

inoperable snubbers is administrative in 

nature and is required for consistency with 

the proposed change to SR 4.7.9. Therefore 

the proposed change does not adversely 

affect plant operations, design functions or 

analyses that verify the capability of systems, 

structures, and components to perform their 

design functions. The consequences of 

accidents previously evaluated are not 

significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve any 

physical alteration of plant equipment. The 

proposed change does not change the method 

by which any safety-related system performs 

its function. As such, no new or different 

types of equipment will be installed, and the 

basic operation of installed equipment is 

unchanged. The methods governing plant 

operation and testing remain consistent with 

current safety analysis assumptions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes ensure snubber 

examination, testing and service life 

monitoring will continue to meet the 

requirements of 10 CFR 50.55a(g) except 

where the NRC has granted specific written 

relief, pursuant to 10 CFR 50.55a(g)(6)(i), or 

authorized alternatives pursuant to 10 CFR 


emcdonald on DSK2BSOYB1PROD with NOTICES
Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 81673 

50.55a(a)(3). Snubbers will continue to be 

demonstrated OPERABLE by performance of 

a program for examination, testing and 

service life monitoring in compliance with 10 

CFR 50.55a or authorized alternatives. The 

proposed change to TS ACTION 3.7.9 for 

inoperable snubbers is administrative in 

nature and is required for consistency with 

the proposed change to SR 4.7.9. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, and with the changes noted 

above in square brackets, it appears that 

the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Vincent 

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

Box 236, Hancocks Bridge, NJ 08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

September 22, 2010, as supplemented 

by letter dated November 22, 2010. 

Description of amendment request: 

The proposed amendment consists of 

changes to the approved fire protection 

program as described in the Wolf Creek 

Generating Station (WCGS) Updated 

Safety Analysis Report (USAR). 

Specifically, amendment proposes a 

deviation from a commitment to certain 

technical requirements of 10 CFR, Part 

50, Appendix R, Section III.L.1, as 

described in Appendix 9.5E of the 

WCGS USAR. The licensee has 

proposed to revise USAR Table 9.5E–1 

to include information on Reactor 

Coolant System process variables not 

maintained within those predicted for a 

loss of normal ac [alternating current] 

power. 

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 design function of structures, systems 

and components (SSCs) are not impacted by 

the proposed change. Evaluation SA–08–006 

Rev. 1 [RETRAN–3D Post-Fire Safe 

Shutdown (PFSSD) Consequence Evaluation 

for a Postulated Control Room Fire] has 

demonstrated that the formation of voids in 

the reactor head for a short time following a 

fire in the control room and spurious 

temporary opening of the pressurizer power 

operated relief valve (PORV) does not result 

in damage to a fission product barrier and 

does not result in a loss of natural circulation 

cooldown. The proposed change does not 

alter or prevent the ability of SSCs from 

performing their intended function to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

Therefore, the probability of any accident 

previously evaluated is not increased. 

Equipment required to mitigate an accident 

remains capable of performing the assumed 

function. 

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 will not alter the 

requirements or function for systems 

required during accident conditions. The 

design function of structures, systems and 

components are not impacted by the 

proposed change. The thermal hydraulic 

analysis of the reactor coolant system 

identified that the process variables are not 

maintained within those predicted for a loss 

of normal ac power, however, the fission 

product boundary integrity is not affected. 

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. 

There will be no effect on the manner in 

which safety limits or limiting safety system 

settings are determined nor will there be any 

effect on those plant systems necessary to 

assure the accomplishment of protection 

functions. There will be no impact on 

departure from nuclear boiling ratio (DNBR) 

limits, heat flux hot channel factor (FQ(Z)) 

limits, nuclear enthalpy rise hot channel 

factor (FNNΔH) limits, peak centerline 

temperature (PCT) limits, peak local power 

density or any other margin of safety. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

November 4, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 5.6.5, 

‘‘CORE OPERATING LIMITS REPORT 

(COLR),’’ to replace the existing large 

break loss-of-coolant accident (LOCA) 

analysis methodology. Specifically, the 

proposed change adds a reference of 

Westinghouse Electric Company’s 

topical report WCAP–16009–P–A, 

‘‘Realistic Large Break LOCA Evaluation 

Methodology Using Automated 

Statistical Treatment of Uncertainty 

Method (ASTRUM),’’ to TS 5.6.5b. 

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 revises TS Section 

5.6.5 to incorporate a new large break LOCA 

analysis methodology. Specifically, the 

proposed change adds WCAP–16009–P–A to 

TS 5.6.5b as a method used for establishing 

core operating limits. 

Accident analyses are not accident 

initiators; therefore, the proposed change 

does not involve a significant increase in the 

probability of an accident. The analyses 

using ASTRUM demonstrated that the 

acceptance criteria in 10 CFR 50.46, 

‘‘Acceptance criteria for emergency core 

cooling systems for lightwater nuclear power 

reactors,’’ were met. Large break LOCA 

analyses performed consistent with the 

methodology in NRC-approved WCAP– 

16009–P–A, including applicable 

assumptions, limitations and conditions, 

demonstrate that 10 CFR 50.46 acceptance 

criteria are met; thus, this change does not 

involve a significant increase in the 

consequences of an accident. No physical 

changes to the plant are associated with the 

proposed change. 

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 previously evaluated? 

Response: No. 

The proposed change revises TS Section 

5.6.5 to incorporate a new large break LOCA 

analysis methodology. Specifically, the 

proposed change adds WCAP–16009–P–A to 

TS 5.6.5b as a method used for establishing 

core operating limits. There are no physical 

changes being made to the plant as a result 


emcdonald on DSK2BSOYB1PROD with NOTICES
81674 Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 

of using the Westinghouse ASTRUM analysis 

methodology in WCAP–16009–P–A for 

performance of the large break LOCA 

analyses. Large break LOCA analyses 

performed consistent with the methodology 

in NRC-approved WCAP–16009–P–A, 

including applicable assumptions, 

limitations and conditions, demonstrate that 

10 CFR 50.46 acceptance criteria are met. No 

new modes of plant operation are being 

introduced. The configuration, operation, 

and accident response of the structures or 

components are unchanged by use of the new 

analysis methodology. Analyses of transient 

events have confirmed that no transient event 

results in a new sequence of events that 

could lead to a new accident scenario. The 

parameters assumed in the analyses are 

within the design limits of existing plant 

equipment. 

In addition, employing the Westinghouse 

ASTRUM large break LOCA analysis 

methodology does not create any new failure 

modes that could lead to a different kind of 

accident. The design of systems remains 

unchanged and no new equipment or 

systems have been installed which could 

potentially introduce new failure modes or 

accident sequences. No changes have been 

made to instrumentation actuation setpoints. 

Adding the reference to WCAP–16009–P–A 

in TS Section 5.6.5b is an administrative 

change that does not create the possibility of 

a new or different kind of accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change revises TS Section 

5.6.5 to incorporate a new large break LOCA 

analysis methodology. Specifically, the 

proposed change adds WCAP–1 6009–P–A to 

TS 5.6.5b as a method used for establishing 

core operating limits. The analyses using 

ASTRUM demonstrated that the applicable 

acceptance criteria in 10 CFR 50.46 are met. 

Margins of safety for large break LOCAs 

include quantitative limits for fuel 

performance established in 10 CFR 50.46. 

These acceptance criteria are not being 

changed by this proposed new methodology. 

Large break LOCA analyses performed 

consistent with the methodology in NRCapproved WCAP–16009–P–A, including 

applicable assumptions, limitations and 

conditions, demonstrate that 10 CFR 50.46 

acceptance criteria are met; thus, this change 

does not involve a significant reduction in a 

margin of safety. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

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. 

NextEra Energy Point Beach, Docket 

Nos. 50–266 and 50–301, Point Beach 

Nuclear Plant (PBNP), Units 1 and 2, 

Manitowoc County, Wisconsin 

Date of amendment request: April 7, 

2009, as supplemented by letters dated 

June 17 (two letters) and December 8 of 

2009; and April 15, July 8, July 28, 

August 24, September 9, September 21, 

October 14, and November 1 of 2010. 

Brief description of amendment 

request: The proposed amendment 

would increase the licensed core power 

level for PBNP Units 1 and 2 from 1540 

to 1800 megawatts thermal. The 

increase in core thermal power will be 

approximately 17 percent over the 

current licensed thermal power level 

and is categorized as an Extended Power 

Uprate. 

Date of publication of individual 

notice in Federal Register: November 

17, 2010 (75 FR 70305). 

Expiration date of individual notice: 

January 18, 2011. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01 F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

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

415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Indiana Michigan Power Company 

(IandM), Docket Nos. 50–315 and 50– 

316, Donald C. Cook Nuclear Plant, 

Units 1 and 2, Berrien County, 

Michigan 

Date of application for amendment: 

September 8, 2010. 

Brief description of amendment: The 

amendments delete the Technical 

Specification requirements related to 

the containment hydrogen recombiners 

and the hydrogen monitors, in 

accordance with Nuclear Energy 

Institute Technical Specification Task 

Force (TSTF) initiative designated as 

TSTF–447. 

Date of issuance: December 14, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 


emcdonald on DSK2BSOYB1PROD with NOTICES
Federal Register / Vol. 75, No. 248 / Tuesday, December 28, 2010 / Notices 81675 

within 120 days from the date of 

issuance. 

Amendment Nos.: 313 (for Unit 1) and 

296 (for Unit 2). 

Facility Operating License Nos. DPR– 

58 and DPR–74: Amendment revised the 

Renewed Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: October 14, 2010 (75 FR 

63209). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 14, 

2010. 

No significant hazards consideration 

comments received: No. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant, Units 1 and 2, Goodhue County, 

Minnesota 

Date of application for amendments: 

November 24, 2009, as supplemented by 

letter dated May 26, 2010. 

Brief description of amendments: 

These amendments revise Technical 

Specification (TS) 4.2.1, ‘‘Fuel 

Assemblies,’’ to add Optimized 

ZIRLOTM as an acceptable fuel rod 

cladding material and add two 

Westinghouse topical reports to the 

analytical methods identified in TS 

5.6.5.b. 

Date of issuance: November 29, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 199, 187. 

Facility Operating License Nos. DPR– 

42 and DPR–60: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

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

The supplemental letter contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 29, 

2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland this 16th day 

of December, 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–32668 Filed 12–27–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0031] 

Notice of Issuance of Regulatory Guide 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice of Issuance and 

Availability of Regulatory Guide 4.16, 

Revision 2, ‘‘Monitoring and Reporting 

Radioactive Materials in Liquid and 

Gaseous Effluents from Nuclear Fuel 

Cycle Facilities.’’ 

FOR FURTHER INFORMATION CONTACT: 

Mekonen M. Bayssie, Regulatory Guide 

Development Branch, Division of 

Engineering, Office of Nuclear 

Regulatory Research, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, telephone (301) 251– 

7489 or e-mail 

Mekonen.Bayssie@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The U.S. Nuclear Regulatory 

Commission (NRC) is issuing a revision 

to an existing guide in the agency’s 

‘‘Regulatory Guide’’ series. This series 

was developed to describe and make 

available to the public information such 

as methods that are acceptable to the 

NRC staff for implementing specific 

parts of the agency’s regulations, 

techniques that the staff uses in 

evaluating specific problems or 

postulated accidents, and data that the 

staff needs in its review of requests for 

licensing actions. In March 2010, 

Revision 2 of Regulatory Guide 4.16, 

‘‘Monitoring and Reporting Radioactive 

Materials in Liquid and Gaseous 

Effluents from Nuclear Fuel Cycle 

Facilities,’’ was published as Draft 

Regulatory Guide, DG–4017, with a 

public comment period of 60 days. This 

guide describes a method that the staff 

of the NRC considers acceptable for the 

development and implementation of 

effluent monitoring programs described 

in license applications and for 

monitoring and reporting effluent data 

by licensees. The guidance is applicable 

to nuclear fuel cycle facilities, with the 

exception of uranium milling facilities 

and nuclear power reactors. The NRC 

has developed other regulatory guides 

applicable to those facilities. 

Revision of this regulatory guide is 

necessary to update references and 

practices and to communicate its 

applicability to the enrichment plants 

which have come under the regulatory 

authority of the NRC since the issuance 

of Revision 1 of the guide. 

II. Further Information 

The staff’s responses to the public 

comments received on DG–4017 are 

located in the NRC’s Agencywide 

Documents Access and Management 

System (ADAMS) under Accession 

Number ML101720322. The regulatory 

analysis may be found in ADAMS under 

Accession No. ML101720311. Electronic 

copies of Regulatory Guide 4.16, 

Revision 2 are available through the 

NRC’s public Web site under 

‘‘Regulatory Guides’’ at http:// 

www.nrc.gov/reading-rm/doccollections/. 

In addition, regulatory guides are 

available for inspection at the NRC’s 

Public Document Room (PDR) located at 

Room O–1F21, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852–2738. The PDR’s 

mailing address is USNRC PDR, 

Washington, DC 20555–0001. The PDR 

can also be reached by telephone at 

(301) 415–4737 or (800) 397–4209, by 

fax at (301) 415–3548, and by e-mail to 

pdr.resource@nrc.gov. 

Regulatory guides are not 

copyrighted, and NRC approval is not 

required to reproduce them. 

Dated at Rockville, Maryland this 15th day 

of December, 2010. 

For the Nuclear Regulatory Commission. 

John N. Ridgely, 

Acting Chief, Regulatory Guide Development 

Branch, Division of Engineering, Office of 

Nuclear Regulatory Research. 

[FR Doc. 2010–32448 Filed 12–27–10; 8:45 am] 

BILLING CODE 7590–01–P 

PENSION BENEFIT GUARANTY 

CORPORATION 

Approval of Exemption From the Bond/ 

Escrow Requirement Relating to the 

Sale of Assets by an Employer Who 

Contributes to a Multiemployer Plan: 

Ricketts Acquisition LLC and the 

Chicago National League Ball Club, 

LLC 

AGENCY: Pension Benefit Guaranty 

Corporation. 

ACTION: Notice of approval. 

SUMMARY: The Pension Benefit Guaranty 

Corporation has granted a request from 

Ricketts Acquisition LLC for an 

exemption from the bond/escrow 

requirement of section 4204(a)(1)(B) of 

the Employee Retirement Income 

Security Act of 1974, as amended, with 

respect to the Major League Baseball 

Players Pension Plan. A notice of the 

request for exemption from the 

requirement was published on 

September 3, 2010. The effect of this 


emcdonald on DSK2BSOYB1PROD with NOTICES