53774 Federal Register / Vol. 74, No. 201 / Tuesday, October 20, 2009 / Notices 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0456] 

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 September 

24, 2009, to October 7, 2009. The last 

biweekly notice was published on 

October 6, 2009 (74 FR 51327). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.92, 

this means that operation of the facility 

in accordance with the proposed 

amendment would not (1) Involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rulemaking and 

Directives Branch (RDB), TWB–05– 

B01M, Division of Administrative 

Services, Office of Administration, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, and 

should cite the publication date and 

page number of this Federal Register 

notice. Written comments may also be 

faxed to the RDB at 301–492–3446. 

Documents may be examined, and/or 

copied for a fee, at the NRC’s Public 

Document Room (PDR), located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area O1F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 


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Federal Register / Vol. 74, No. 201 / Tuesday, October 20, 2009 / Notices 53775 

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, 

which the NRC promulgated in August 

28, 2007 (72 FR 49139). The E-Filing 

process requires participants to submit 

and serve all adjudicatory documents 

over the internet, or in some cases to 

mail copies on electronic storage media. 

Participants may not submit paper 

copies of their filings unless they seek 

an exemption in accordance with the 

procedures described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

(10) days prior to the filing deadline, the 

petitioner/requestor should contact the 

Office of the Secretary by e-mail at 

hearing.docket@nrc.gov, or by calling 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

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

petitioner/requestor will need to 

download the Workplace Forms 

ViewerTM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

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

Information about applying for a digital 

ID certificate is available on NRC’s 

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

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

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it can then submit a request for 

hearing or petition for leave to 

intervene. Submissions should be in 

Portable Document Format (PDF) in 

accordance with NRC guidance 

available on the NRC public Web site at 

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

complete at the time the filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

the EIE system no later than 11:59 p.m. 

Eastern Time on the due date. Upon 

receipt of a transmission, the E-Filing 

system time-stamps the document and 

sends the submitter an e-mail notice 

confirming receipt of the document. The 

EIE system also distributes an e-mail 

notice that provides access to the 

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the agency’s adjudicatory e-filing system 

may seek assistance through the 

‘‘Contact Us’’ link located on the NRC 

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

NRC Meta-System Help Desk, which is 

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

Eastern Time, Monday through Friday, 

excluding government holidays. The 

Meta-System Help Desk can be 

contacted by telephone at 1–866–672– 

7640 or by e-mail at 

MSHD.Resource@nrc.gov. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

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

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

the Atomic Safety and Licensing Board 

that the request and/or petition should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

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

unless excluded pursuant to an order of 

the Commission, an Atomic Safety and 

Licensing Board, or a Presiding Officer. 

Participants are requested not to include 

personal privacy information, such as 

social security numbers, home 

addresses, or home phone numbers in 

their filings, unless an NRC regulation 

or other law requires submission of such 

information. With respect to 

copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submissions. 

For further details with respect to this 

license amendment application, see the 

application for amendment which is 

available for public inspection at the 

Commission’s PDR, located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. Publicly 

available records will be accessible from 

the ADAMS Public Electronic Reading 

Room on the Internet at the NRC Web 

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

adams.html. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 


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53776 Federal Register / Vol. 74, No. 201 / Tuesday, October 20, 2009 / Notices 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station (RBS), Unit 1, 

West Feliciana Parish, Louisiana 

Date of amendment request: August 

10, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the RBS Technical Specifications (TSs) 

to support operation with 24-month fuel 

cycles. Specifically, the change 

addresses certain TS Surveillance 

Requirement (SR) frequencies that are 

specified as ‘‘18 months’’ by revising 

them to ‘‘24 months’’ in accordance 

with the guidance of U.S. Nuclear 

Regulatory Commission (NRC) Generic 

Letter 91–04, ‘‘Changes in Technical 

Specification Surveillance Intervals to 

Accommodate a 24-Month Fuel Cycle,’’ 

dated April 2, 1991. 

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 TS changes involve a change 

in the surveillance testing intervals and 

allowable values to facilitate a change in the 

operating cycle length. The proposed TS 

changes do not physically impact the plant. 

The proposed TS changes do not degrade the 

performance of, or increase the challenges to, 

any safety systems assumed to function in 

the accident analysis. The proposed TS 

changes do not impact the usefulness of the 

SRs in evaluating the operability of required 

systems and components, or the way in 

which the surveillances are performed. In 

addition, the frequency of surveillance 

testing is not considered an initiator of any 

analyzed accident, nor does a revision to the 

frequency introduce any accident initiators. 

The specific value of the allowable value is 

not considered an initiator of any analyzed 

accident. Therefore, the proposed change 

does not involve a significant increase in the 

probability of an accident previously 

evaluated. 

The consequences of a previously 

evaluated accident are not significantly 

increased. The proposed change does not 

affect the performance of any equipment 

credited to mitigate the radiological 

consequences of an accident. Evaluation of 

the proposed TS changes demonstrated that 

the availability of credited equipment is not 

significantly affected because of other more 

frequent testing that is performed, the 

availability of redundant systems and 

equipment, and the high reliability of the 

equipment. Historical review of surveillance 

test results and associated maintenance 

records did not find evidence of failures that 

would invalidate the above conclusions. 

The allowable values have been developed 

in accordance with [NRC Regulatory Guide] 

1.105, ‘‘Instrument Setpoints,’’ to ensure that 

the design and safety analysis limits are 

satisfied. The methodology used for the 

development of the allowable values ensures 

the affected instrumentation remains capable 

of mitigating design basis events as described 

in the safety analyses and that the results and 

radiological consequences described in the 

safety analyses remain bounding. Therefore, 

the proposed change does not alter the ability 

to detect and mitigate events and, as such, 

does not involve a significant increase in the 

consequences of an accident previously 

evaluated. 

Standby Liquid Control System 

The proposed change in required weight of 

Boron-10 in [standby liquid control (SLC)] 

does not physically impact the plant, nor 

does it degrade the performance of, or 

increase the challenges to, any safety systems 

assumed to function in the accident analysis. 

The consequences of a previously evaluated 

accident are not increased. The proposed 

change does not affect the performance of 

any equipment credited to mitigate the 

radiological consequences of an accident. 

Evaluation of the proposed TS changes 

demonstrated that the availability of credited 

equipment is not affected. Therefore, the 

proposed change does not alter the ability to 

detect and mitigate events and, as such, does 

not involve a significant increase in the 

consequences of an accident previously 

evaluated. 

Loss of Power Instrumentation 

A change to the Allowable Values (AVs) is 

proposed for Table 3.3.8.1–1, Item 1.c and 

Item 2.c. The proposed change is the result 

of application of the RBS Instrument Setpoint 

Methodology using plant-specific drift values 

and incorporating margins available based on 

a revised off-site reliability study. 

Application of this methodology results in 

AVs that more accurately reflect total device 

accuracy, as well as that of test equipment 

and calculated drift between surveillances. 

The proposed change will not result in any 

hardware changes. The instrumentation is 

not assumed to be an initiator of any 

analyzed event. Existing operating margin 

between plant conditions and actual plant 

setpoints is not significantly reduced due to 

the proposed changes. The role of the 

instrumentation is in mitigating and thereby, 

limiting the consequences of accidents. 

The AVs were developed to ensure the 

design and safety analysis limits are satisfied. 

The methodology used for the development 

of the AVs ensures that: (1) The affected 

instrumentation remains capable of 

mitigating design basis events as described in 

the safety analysis, and (2) the results and 

radiological consequences described in the 

safety analysis remain bounding. 

Additionally, the proposed change does not 

alter the plant’s ability to detect and mitigate 

events. Therefore, this change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The change in the degraded voltage 

protection voltage AVs allows the protection 

scheme to function as originally designed. 

The proposed allowable values ensure that 

the Class 1E distribution system remains 

connected to the offsite power system when 

adequate offsite voltage is available and 

motor starting transients are considered. 

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 TS changes involve a change 

in the surveillance testing intervals and 

allowable values to facilitate a change in the 

operating cycle length. The proposed TS 

changes do not introduce any failure 

mechanisms of a different type than those 

previously evaluated, since there are no 

physical changes being made to the facility. 

No new or different equipment is being 

installed. No installed equipment is being 

operated in a different manner. As a result, 

no new failure modes are being introduced. 

The way surveillance tests are performed 

remains unchanged. A historical review of 

surveillance test results and associated 

maintenance records indicated there was no 

evidence of any failures that would 

invalidate the above conclusions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Standby Liquid Control System 

The proposed change to the required 

weight of Boron-10 in SLC does not 

introduce any failure mechanisms of a 

different type than those previously 

evaluated, since there are no physical 

changes being made to the facility. No new 

or different equipment is being installed. No 

installed equipment is being operated in a 

different manner. As a result, no new failure 

modes are being introduced. The way 

surveillance tests are performed remains 

unchanged. A historical review of 

surveillance test results and associated 

maintenance records indicated there was no 

evidence of any failures that would 

invalidate the above conclusions. 

Loss of Power Instrumentation 

The proposed change in AVs is the result 

of application of the Instrument Setpoint 

Methodology using plant-specific drift values 

and does not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. This is based 

upon the fact that the method and manner of 

plant operation are unchanged. 

The use of the proposed AVs does not 

impact safe operation of the plant in that the 

safety analysis limits are maintained. The 

proposed change in AVs involves no system 

additions. The AVs are revised to ensure the 

affected instrumentation remains capable of 

mitigating accidents and transients. Plant 

equipment will not be operated in a manner 

different from previous operation, except that 

setpoints may be changed. No additional 

failure mechanisms are introduced as a result 

of the changes to the allowable values. Since 

operational methods remain unchanged and 


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the operating parameters were evaluated to 

maintain the plant within existing design 

basis criteria, no different type of failure or 

accident is created. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed TS changes involve a change 

in the surveillance testing intervals and 

allowable values to facilitate a change in the 

operating cycle length. The impact of these 

changes on system availability is not 

significant, based on other more frequent 

testing that is performed, the existence of 

redundant systems and equipment, and 

overall system reliability. Evaluations have 

shown there is no evidence of time 

dependent failures that would impact the 

availability of the systems. The proposed 

changes do not significantly impact the 

condition or performance of structures, 

systems, and components relied upon for 

accident mitigation. The proposed changes in 

TS instrumentation allowable values are the 

result of application of the RBS setpoint 

methodology using plant specific drift 

values. The revised allowable values more 

accurately reflect total instrumentation loop 

accuracy including drift while continuing to 

protect any assumed analytical limit. The 

proposed changes do not result in any 

hardware changes or in any changes to the 

analytical limits assumed in accident 

analyses. Existing operating margin between 

plant conditions and actual plant setpoints is 

not significantly reduced due to these 

changes. The proposed changes do not 

significantly impact any safety analysis 

assumptions or results. 

Standby Liquid Control System 

The proposed change in required weight of 

Boron-10 in SLC is to facilitate a change in 

the operating cycle length. The proposed 

change does not result in any hardware 

changes or in any changes to the analytical 

limits assumed in accident analyses. Existing 

operating margin between plant conditions 

and actual plant setpoints is not reduced due 

to this change. The proposed change does not 

impact any safety analysis assumptions or 

results. Therefore, the proposed change does 

not involve a significant reduction in a 

margin of safety. 

Loss of Power Instrumentation 

The proposed protection voltage AVs are 

low enough to prevent inadvertent power 

supply transfer, but high enough to ensure 

that sufficient voltage is available to the 

required equipment. The proposed change 

does not involve a reduction in a margin of 

safety. The proposed change was developed 

using a methodology to ensure safety analysis 

limits are not exceeded. As such, this 

proposed 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: Terence A. 

Burke, Associate General Counsel— 

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Nuclear 

Plant, Van Buren County, Michigan 

Date of amendment request: August 

25, 2009. 

Description of amendment request: 

The proposed amendment would allow 

for a one-time extension to the ten-year 

frequency for the next Palisades Nuclear 

Plant (PNP) containment Type A 

integrated leak rate test (ILRT) that is 

required by Technical Specification (TS) 

5.5.14. The proposed change would 

permit the existing ILRT frequency to be 

extended from ten years to 

approximately 11.25 years. 

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 exemption involves a onetime extension to the current interval for 

Type A containment testing. The current test 

interval of 120 months (10 years) would be 

extended on a one-time basis to no longer 

than approximately 135 months from the last 

Type A test. The proposed extension does 

not involve either a physical change to the 

plant or a change in the manner in which the 

plant is operated or controlled. The 

containment is designed to provide an 

essentially leak tight barrier against the 

uncontrolled release of radioactivity to the 

environment for postulated accidents. As 

such, the containment and the testing 

requirements invoked to periodically 

demonstrate the integrity of the containment 

exist to ensure the plant’s ability to mitigate 

the consequences of an accident, and do not 

involve the prevention or identification of 

any precursors of an accident. Therefore, this 

proposed extension does not involve a 

significant increase in the probability of an 

accident previously evaluated. 

This proposed extension is for the Type A 

containment leak rate tests only. The Type B 

and C containment leak rate tests would 

continue to be performed at the frequency 

currently required by the PNP TS. As 

documented in NUREG 1493, Type B and C 

tests have identified a very large percentage 

of containment leakage paths and that the 

percentage of containment leakage paths that 

are detected only by Type A testing is very 

small. The PNP Type A test history supports 

this conclusion. 

The integrity of the containment is subject 

to two types of failure mechanisms that can 

be categorized as (1) activity based and (2) 

time based. Activity based failure 

mechanisms are defined as degradation due 

to system and/or component modifications or 

maintenance. Local leak rate test 

requirements and administrative controls 

such as configuration management and 

procedural requirements for system 

restoration ensure that containment integrity 

is not degraded by plant modifications or 

maintenance activities. The design and 

construction requirements of the 

containment combined with the containment 

inspections performed in accordance with 

ASME Section XI, the Maintenance Rule, and 

TS requirements serve to provide a high 

degree of assurance that the containment 

would not degrade in a manner that is 

detectable only by a Type A test. Based on 

the above, the proposed extension does not 

involve a significant increase in the 

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 revision to the TS involves 

a one-time extension to the current interval 

for Type A containment testing. The 

containment and the testing requirements 

invoked to periodically demonstrate the 

integrity of the containment exist to ensure 

the plant’s ability to mitigate the 

consequences of an accident and do not 

involve the prevention or identification of 

any precursors of an accident. The proposed 

TS change does not involve a physical 

change to the plant or the manner in which 

the plant is operated or controlled. Therefore, 

the proposed TS 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? 

Response: No. 

The proposed change to the TS involves a 

one-time extension to the current interval for 

Type A containment testing. The proposed 

TS change does not involve a physical 

change to the plant or a change in the manner 

in which the plant is operated or controlled. 

The specific requirements and conditions of 

the TS Containment Leak Rate Testing 

Program exist to ensure that the degree of 

containment structural integrity and leaktightness that is considered in the plant 

safety analysis is maintained. The overall 

containment leak rate limit specified by TS 

is maintained. The proposed change involves 

only the extension of the interval between 

Type A containment leak rate tests. The 

proposed surveillance interval extension is 

bounded by the 15-month extension 

currently authorized within NEI 94–01, 

Revision 0. Type B and C containment leak 


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rate tests would continue to be performed at 

the frequency currently required by TS. 

Industry experience supports the conclusion 

that Type B and C testing detects a large 

percentage of containment leakage paths and 

that the percentage of containment leakage 

paths that are detected only by Type A 

testing is small. The containment inspections 

performed in accordance with ASME Section 

XI and the Maintenance Rule serve to 

provide a high degree of assurance that the 

containment would not degrade in a manner 

that is detectable only by Type A testing. The 

combination of these factors ensures that the 

margin of safety in the plant safety analysis 

is maintained. The design, operation, testing 

methods and acceptance criteria for Type A, 

B, and C containment leakage tests specified 

in applicable codes and standards would 

continue to be met, with the acceptance of 

this proposed change, since these are not 

affected by changes to the Type A test 

interval. Therefore, the proposed TS 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 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Ave., White Plains, NY 10601. 

NRC Acting Branch Chief: Peter Tam. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: August 

26, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) Section 

6.5 that governs administrative controls 

of High Radiation Areas (HRA) to 

incorporate the HRA administrative 

controls contained within the Standard 

Technical Specifications, NUREG–1433, 

Revision 3. 

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 operation of Vermont Yankee 

Nuclear Power Station (VY) in accordance 

with the proposed amendment will not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed amendment does not impact 

the operability of any structure, system or 

component that affects the probability of an 

accident or that supports mitigation of an 

accident previously evaluated. The proposed 

amendment does not affect reactor operations 

or accident analysis and has no radiological 

consequences. The operability requirements 

for accident mitigation systems remain 

consistent with the licensing and design 

basis. Therefore, the proposed amendment 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The operation of VY in accordance with 

the proposed amendment will not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed amendment does not change 

the design or function of any component or 

system. No new modes of failure or initiating 

events are being introduced. Therefore, 

operation of VY in accordance with the 

proposed amendment will not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The operation of VY in accordance with 

the proposed amendment will not involve a 

significant reduction in a margin of safety. 

The proposed amendment does not change 

the design or function of any component or 

system. The proposed amendment does not 

involve any safety limits, safety settings or 

safety margins. The TS administrative access 

controls for high radiation areas are being 

replaced with those contained in section 5.7 

of NUREG–1433 to provide additional 

requirements and options for the control of 

these areas. 

Therefore, operation of VY in accordance 

with the proposed amendment will not 

involve a significant reduction in the margin 

to 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., 400 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy Salgado. 

Nine Mile Point Nuclear Station, LLC 

(NMPNS) Docket No. 50–410, Nine Mile 

Point Nuclear Station Unit No. 2 (NMP 

2), Oswego County, New York 

Date of amendment request: May 27, 

2009, as supplemented on August 28, 

2009. 

Description of amendment request: 

The proposed amendment requests an 

increase in the maximum steady-state 

power level at NMP2 from 3467 

megawatts thermal (MWt) to 3988 MWt. 

This represents a 15-percent increase 

over the current licensed thermal 

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? 

No, the increase in power level discussed 

herein will not significantly increase the 

probability or consequences of an accident 

previously evaluated. 

The proposed change will increase NMP2’s 

authorized maximum power level from the 

current licensed thermal power (CLTP) level 

of 3467 megawatts thermal (MWt) to 3988 

MWt. In support of this Constant Pressure 

Extended Power Uprate (CPPU), a 

comprehensive evaluation was performed for 

nuclear steam supply system (NSSS) and 

balance of plant (BOP) systems, structures, 

components, and analyses that could be 

affected by this change. The effect of 

increasing the maximum power level from 

the CLTP of 3467 MWt to 3988 MWt on the 

NMP2 licensing and design bases was 

evaluated. The result of this evaluation is 

that all plant components, as modified, will 

continue to be capable of performing their 

design function at an uprated core power of 

3988 MWt. In addition, an evaluation of the 

accident analyses concludes that applicable 

analysis acceptance criteria continue to be 

met. Power level is an input assumption to 

the equipment design and accident analyses, 

but it is not an initiator for any transient or 

accident. Therefore, no accident initiators are 

affected by this uprate and no challenges to 

any plant safety barriers are created by this 

change. 

Therefore, operation of the facility in 

accordance with the proposed change does 

not involve a significant increase in the 

probability of an accident previously 

evaluated. 

This change does not affect the release 

paths, the frequency of release, or the source 

term for release for any accidents previously 

evaluated in the Updated Safety Analysis 

Report (USAR). Structures, systems, and 

components (SSC) required to mitigate 

transients remain capable of performing their 

design functions, and thus were found 

acceptable. The source terms used to assess 

radiological consequences have been 

reviewed and determined to bound operation 

at the uprated condition. The results of EPU 

[extended power uprate] accident evaluations 

do not exceed the U. S. Nuclear Regulatory 

Commission (NRC) approved acceptance 

limits. 

The spectrum of postulated accidents and 

transients has been investigated and are 

shown to meet the regulatory criteria to 

which NMP2 is currently licensed. In the 

area of fuel and core design, the Safety Limit 

Minimum Critical Power ratio (SLMCPR) and 

other applicable Specified Acceptable Fuel 

Design Limits (SAFDLS) are still met. 

Continued compliance with the SLMCPR and 

other SAFDLs is confirmed on a cycle 

specific basis consistent with criteria 

accepted by the NRC. 

Challenges to the reactor coolant pressure 

boundary were evaluated at EPU conditions 


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(pressure, temperature, flow, and radiation) 

and found to meet the acceptance criteria for 

allowable stresses. Adequate overpressure 

margin is maintained. 

Challenges to the containment have been 

evaluated and the containment and its 

associated cooling system continue to meet 

applicable regulatory requirements. The 

increase in the calculated post Loss of 

Coolant Accident (LOCA) suppression pool 

temperature above the current peak 

temperature was evaluated and determined 

to be acceptable. 

Radiological release events (accidents) 

have been evaluated and shown to meet the 

requirements of 10 CFR 50.67. 

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? 

No, the increase in power level discussed 

herein will not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

The proposed change will increase NMP2’s 

authorized maximum power level from the 

CLTP level of 3467 MWt to 3988 MWt. 

Equipment that could be affected by EPU has 

been evaluated. No new operating mode, 

safety-related equipment lineup, accident 

scenario, or equipment failure mode was 

identified. The full spectrum of accident 

considerations has been evaluated and no 

new or different kind of accident has been 

identified. This Constant Pressure Extended 

Power Uprate utilizes a standard evaluation 

methodology applied to known technology 

employed within the range of current or 

modified plant capabilities. As such, the 

plant safety-related equipment continues to 

operate in accordance with regulatory 

criteria. Evaluations were performed using 

NRC approved codes, standards and 

methods. No new accidents or event 

precursors have been identified. 

All structures, systems and components 

previously required for the mitigation of a 

transient remain capable of fulfilling their 

intended design functions. The proposed 

changes do not adversely affect safety-related 

systems or components and do not challenge 

the performance or integrity of any safetyrelated system. This change does not 

adversely affect any current system interfaces 

or create any new interfaces that could result 

in an accident or malfunction of a different 

kind than was previously evaluated. 

Operating at a core power level of 3988 MWt 

does not create any new accident initiators or 

precursors. 

Therefore, the proposed changes do 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? 

No, the increase in power level discussed 

herein will not involve a significant 

reduction in a margin of safety. 

Comprehensive analyses of the proposed 

changes have concluded that relevant design 

and safety acceptance criteria will be met 

without a significant reduction in margins of 

safety. The analyses supporting EPU have 

demonstrated that the NMP2 SSCs are 

capable of safely performing at EPU 

conditions. The analyses identified and 

defined the major input parameters to the 

NSSS, analyzed NSSS design transients, and 

evaluated the capabilities of the NSSS fluid 

systems, NSSS/BOP interfaces, NSSS control 

systems, and NSSS and BOP components, as 

appropriate. Radiological consequences of 

design basis events remain within regulatory 

limits and are not increased significantly. 

The analyses confirmed that NSSS and BOP 

SSCs are capable, some with modifications, 

of achieving EPU conditions without 

significant reduction in margins of safety. 

Analyses have shown that the integrity of 

primary fission product barriers will not be 

significantly affected as a result of the power 

increase. Calculated loads on SSCs important 

to safety have been shown to remain within 

design allowables under EPU conditions for 

all design basis event categories. Plant 

response to transients and accidents do not 

result in exceeding acceptance criteria. As 

appropriate, the evaluations that demonstrate 

acceptability of EPU have been performed 

using methods that have either been 

reviewed and approved by the NRC staff, or 

that are in compliance with regulatory review 

guidance and standards established for 

maintaining adequate margins of safety. 

These evaluations demonstrate that there are 

no significant reductions in the margins of 

safety. 

Maximum power level is one of the 

inherent inputs that determine the safe 

operating range defined by the accident 

analyses. The Technical Specifications 

ensure that NMP2 is operated within the 

bounds of the inputs and assumptions used 

in the accident analyses. The acceptance 

criteria for the accident analyses are 

conservative with respect to the operating 

conditions defined by the Technical 

Specifications. The engineering reviews 

performed for the constant pressure extended 

power uprate confirm that the accident 

analyses criteria are met at the revised 

maximum allowable thermal power level of 

3988 MWt, as well as at the rated thermal 

power (RTP) levels specified in the Facility 

Operating License and Technical 

Specifications. Therefore, the adequacy of the 

revised Facility Operating Licenses and 

Technical Specifications to maintain the 

plant in a safe operating range is also 

confirmed, and the increase in maximum 

allowable power level does not involve a 

significant decrease in a margin of safety. 

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: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Branch Chief: Nancy L. Salgado. 

Nine Mile Point Nuclear Station, LLC 

(NMPNS) Docket No. 50–410, Nine Mile 

Point Nuclear Station Unit No. 2 

(NMP2), Oswego County, New York 

Date of amendment request: June 29, 

2009, as supplemented on August 13, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

the NMP2 Technical Specification (TS) 

5.5.12 by replacing the reference to 

Regulatory Guide (RG) 1.163 with a 

reference to Nuclear Energy Institute 

(NEI) Topical Report NEI 94–01, 

Revision 2–A, as the implementation 

document used by NMPNS to develop 

the NMP2 performance-based leakage 

testing program in accordance with 

Option B of Title 10 of the Code of 

Federal Regulations (10 CFR) Part 50. 

The proposed amendment would allow 

the next primary containment integrated 

leak rate test (ILRT) to be performed 

within 15 years from the last ILRT as 

opposed to the current 10-year interval, 

and would allow successive ILRTs to be 

performed at 15-year intervals. 

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 involves 

changes to the NMP2 10 CFR 50 Appendix 

J Testing Program Plan. The proposed 

amendment does not involve a physical 

change to the plant or a change in the manner 

in which the plant is operated or controlled. 

The primary containment function is to 

provide an essentially leak tight barrier 

against the uncontrolled release of 

radioactivity to the environment for 

postulated accidents. As such, the 

containment itself and the testing 

requirements to periodically demonstrate the 

integrity of the containment exist to ensure 

the plant’s ability to mitigate the 

consequences of an accident, and do not 

involve any accident precursors or initiators. 

Therefore, the probability of occurrence of an 

accident previously evaluated is not 

significantly increased by the proposed 

amendment. 

The proposed amendment adopts the NRCaccepted guidelines of NEI 94–01, Revision 2, 

for development of the NMP2 performancebased leakage testing program. 

Implementation of these guidelines continues 

to provide adequate assurance that during 

design basis accidents, the primary 

containment and its components will limit 


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leakage rates to less [then] the values 

assumed in the plant safety analyses. The 

potential consequences of extending the ILRT 

interval from 10 years to 15 years have been 

evaluated by analyzing the resulting changes 

in risk. The increase in risk in terms of 

person-rem per year within 50 miles 

resulting from design basis accidents was 

estimated to be acceptably small, and the 

increase in the large early release frequency 

resulting from the proposed change was 

determined to be within the guidelines 

published in NRC RG 1.174. Additionally, 

the proposed change maintains defense-indepth by preserving a reasonable balance 

among prevention of core damage, 

prevention of containment failure, and 

consequence mitigation. NMPNS has 

determined that the increase in conditional 

containment failure probability due to the 

proposed change would be very small. 

Therefore, it is concluded that the proposed 

amendment does not significantly increase 

the 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 adopts the NRCaccepted guidelines of NEI–94–01, Revision 

2, for development of the NMP2 

performance-based leakage testing program, 

and establishes a 15 year interval for the 

performance of the primary containment 

ILRT. The containment and the testing 

requirements to periodically demonstrate the 

integrity of the containment exist to ensure 

the plant’s ability to mitigate the 

consequences of an accident, and do not 

involve any accident precursors and 

initiators. The proposed change does not 

involve a physical change to the plant (i.e., 

no new or different type of equipment will 

be installed) or a change to the manner in 

which the plant is operated or controlled. 

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 amendment adopts the NRCaccepted guidelines of NEI–94–01, Revision 

2, for development of the NMP2 

performance-based leakage testing program, 

and establishes a 15 year interval for the 

performance of the primary containment 

ILRT. The amendment does not alter the 

manner in which safety limits, limiting safety 

system setpoints, or limiting conditions for 

operation are determined. The specific 

requirements and conditions of the 10 CFR 

50 Appendix J Testing Program Plan, as 

defined in the TS, ensure that the degree of 

primary containment structural integrity and 

leak-tightness that is considered in the plant 

safety analyses is maintained. The overall 

containment leakage rate limit specified by 

the TS is maintained, and the Type A, B, and 

C containment leakage tests will continue to 

be performed at the frequencies established 

in accordance with the NRC-accepted 

guidelines of NEI 94–01, Revision 2. 

Containment inspections performed in 

accordance with other plant programs serve 

to provide a high degree of assurance that the 

containment will not degrade in a manner 

that is detectable only by an ILRT. In 

addition, the on-line containment monitoring 

capability that is inherent to inerted boiling 

water reactor containments allows for the 

detection of gross containment leakage that 

may develop during power operation. This 

combination of factors ensures that evidence 

of containment structural degradation is 

identified in a timely manner. Furthermore, 

a risk assessment using the current NMP2 

Probabilistic Risk Assessment model 

concluded that extending the ILRT test 

interval from 10 years to 15 years results in 

a very small change to the NMP2 risk profile. 

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: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Branch Chief: Nancy L. Salgado. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: 

September 15, 2009. 

Description of amendment request: 

The proposed amendment revises 

Technical Specification (TS) 3.3.2, in 

Appendix A to Facility Operating 

License Nos. NPF–2 and NPF–8 for the 

Joseph M. Farley Nuclear Plant Units 1 

and 2, respectively. P–11 is an 

engineered safety feature actuation 

system (ESFAS) permissive/interlock 

which permits normal unit cooldown 

and depressurization without actuation 

of safety injection (SI) from low 

pressurizer pressure. P–12 is an ESFAS 

permissive/interlock which permits 

normal unit cooldown and 

depressurization without actuation of SI 

and main steam line isolation on the 

condition of low steam line pressure. 

Both P–11 and P–12 circuits use input 

from three protection channels. The 

current wording of Condition K in TS 

3.3.2 states, ‘‘Two channels inoperable.’’ 

As a result, Condition K does not 

explicitly address the possible 

conditions of one channel or three 

channels inoperable, possibly creating a 

literal compliance issue. The proposed 

Condition K change from ‘‘Two 

channels inoperable’’ to ‘‘One or more 

channels inoperable’’ will resolve the 

current literal compliance issue. The 

change does not alter the current 

Condition K required action, it simply 

clarifies that the required action must be 

performed for one, two, or three P–11 or 

P–12 channels inoperable. In addition, 

an editorial change is proposed for TS 

5.6.8 to correct the citation of a 

condition requiring a report for the postaccident monitoring instrumentation. 

The current TS 5.6.8 text states, ‘‘When 

a report is required by Condition B or 

G of LCO [limiting conditions for 

operation] 3.3.3. * * *’’ The citation of 

Condition B is correct while Condition 

G does not currently exist for LCO 3.3.3; 

instead TS 5.6.8 should cite Condition 

F. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (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 TS 3.3.2 does not 

significantly increase the probability or 

consequences of an accident previously 

evaluated in the FSAR. These interlocks do 

not directly initiate an accident. The 

consequences of accidents previously 

evaluated in the FSAR are not adversely 

affected by these changes because the 

changes are made to reflect the Improved 

Standard Technical Specifications and the 

interlocks are verified to be in the required 

state for the unit condition. 

The proposed change to TS 5.6.8 corrects 

an editorial error and therefore does not 

significantly increase the probability or 

consequences of a previously evaluated 

accident. 

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 TS 3.3.2 does not 

create the possibility of a new or different 

kind of accident than any accident already 

evaluated in the FSAR. No new accident 

scenario, failure mechanisms, or limiting 

single failures are introduced as a result of 

the proposed change. The proposed TS 3.3.2 

change does not challenge the performance 

or integrity of any safety-related systems. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

analyzed. 

The proposed change to TS 5.6.8 corrects 

an editorial error and therefore does not 

create the possibility of a new or different 

kind of accident from any accident 

previously analyzed. 


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3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change to TS 3.3.2 does not 

involve a significant reduction in a margin of 

safety. The proposed change is made to 

accurately reflect the format of the Improved 

Standard Technical Specifications. The 

actuation setpoints specified by the 

Technical Specifications and safety analysis 

limits assumed in the accident analysis are 

unchanged. The margin of safety associated 

with these trip setpoints and the safety 

analysis acceptance criteria is unchanged. 

Therefore, the proposed change to TS 3.3.2 

will not significantly reduce the margin of 

safety as defined in the Technical 

Specifications. 

The proposed change to TS 5.6.8 corrects 

an editorial error and therefore involves no 

significant reduction in a margin of safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 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: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Branch Chief: Jon H. Thompson, 

Acting. 

Tennessee Valley Authority (TVA), 

Docket Nos. 50–259, 50–260 and 50– 

296, Browns Ferry Nuclear Plant, Units 

1, 2 and 3, Limestone County, Alabama 

Date of amendment request: July 27, 

2009 (TS–465). 

Description of amendment request: 

The proposed change is to eliminate 

Technical Specification (TS) 

surveillance requirement (SR) 3.6.1.3.11 

and the requirement to perform water 

leak rate testing on the listed 

containment isolation valves. More 

specifically, the proposed change 

eliminates water local leak rate testing 

of valves in the Containment Leak Rate 

Program that are being tested to verify 

the combined leakage rate is within the 

limit that ensures the suppression pool 

level is sufficient to keep lines that 

terminate below the water level for at 

least 30 days without additional makeup. 

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. 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

This proposal does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed change to the scope of water 

leak rate testing for the subject valves does 

not affect the probability of the design basis 

accidents. The valves will continue to be 

maintained in an operable state, and in their 

current design configuration. There is no 

correlation between the scope of the water 

leak rate testing and accident probability. 

TVA reviewed the postulated 

consequences of design basis events on 

primary containment isolation under the 

proposed change. The primary containment 

structure, including access openings, 

penetrations and the containment heat 

removal system, is designed so that the 

containment structure and its internal 

compartments can withstand, without 

exceeding the design leakage rate (2.0% per 

day), the peak accident pressure and 

temperature that could occur during any 

postulated LOCA [loss-of-coolant accident]. 

For the purposes of considering the 

consequences of LOCAs under the proposed 

change, a single active failure of a CIV 

[containment isolation valve] or a passive 

failure of the closed system were reviewed, 

within the limits of the existing licensing 

basis. Under the existing licensing basis, a 

pipe rupture of seismically qualified ECCS 

[emergency core cooling system] piping does 

not have to be assumed concurrent with the 

LOCA, except if it is a consequence of the 

LOCA. Consequential failures can be 

eliminated, since a LOCA inside containment 

is separated from the ECCS piping by the 

containment structure. Consequential failures 

of the ECCS piping from LOCA’s outside 

containment are outside the Appendix J 

design considerations, although they are 

adequately addressed through the 

redundancy and separation of the ECCS 

design. A single active failure of the CIV, 

under the LOCA condition, can be 

accommodated since the closed and filled 

system piping and the suppression pool 

water inventory remain as the leakage 

barriers. The ECCS passive failure criterion 

does require consideration of system leaks, 

but not pipe breaks, beyond the initiating 

LOCA. Pipe leakage, equivalent to the 

leakage from a valve or pump seal failure, 

should be considered at 24 hours or greater 

post-LOCA. The capability to make-up 

inventory to the suppression pool is adequate 

to ensure that postulated seat leakage and 

pipe leakage does not result in a condition 

that jeopardizes pool level. Make-up 

capability exists to the suppression pool. 

Actions to make-up to the suppression pool 

are delineated in Emergency Operating 

Instructions. 

Therefore, the proposal to eliminate the 

subject water leak rate tests 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? 

This proposal does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The acceptability of the proposed change 

to the scope of water leak rate testing for the 

subject valves is based on maintaining the 

existing barriers to primary containment 

leakage, and ensuring that the suppression 

pool level is assured for 30 days during all 

design basis, post-accident modes of 

operation. By meeting these dual objectives, 

the plant response to the design basis events 

will be unchanged, and no new accident 

scenarios will be encountered. These two 

objectives are related, in that, the 

suppression pool inventory creates a passive 

barrier to primary containment atmospheric 

leakage for valves associated with 

penetrations which are located below the 

minimum water level of the pool. 

The proposed Technical Specification 

change does not alter the configuration of the 

subject containment isolation valves or their 

associated systems. The valves will continue 

to be tested and maintained to ensure their 

operability. The subject valves are all 

isolation valves associated with lines that 

penetrate the primary containment. For 

closed system valves, the redundant isolation 

boundary for each of the affected valves is 

the closed system associated with the valve. 

The closed system piping is verified via a 10 

CFR 50 Appendix J Type A test. The integrity 

of the closed systems is also monitored and 

controlled via Technical Specification 5.5.2, 

‘‘Primary Coolant Sources Outside 

Containment.’’ 

The subject valves may be open, or change 

state, post-accident to support the design 

function of their associated ECCS systems 

(HPCI [high-pressure coolant injection], Core 

Spray, RHR [residual heat removal]), RCIC 

[reactor core isolation coolant] or RHR 

Sampling using the Post Accident Sampling 

System. The subject valves function as 

system valves during the periods when they 

are open or in an intermediate state, not as 

containment isolation valves. Reliance is 

placed on the suppression pool seal and the 

closed system piping to maintain the barrier 

between primary and secondary containment 

atmospheres. 

Therefore, with the valve configuration and 

closed systems configuration unaffected by 

the proposed change, the existing barriers to 

primary containment atmospheric leakage are 

maintained, so long as the suppression pool 

level is ensured. 

The suppression pool is designed and 

operated so that it is filled with water in 

accordance with Technical Specifications 

3.6.2.2, ‘‘Suppression Pool Water Level,’’ and 

the associated Bases. As such, the supply of 

water in the suppression pool is assured for 

30 days during all design basis, post-accident 

modes of operation. Water leak rate testing 

has historically been performed on valves 

associated with lines that connect to the 

suppression pool. The acceptance criteria for 

combined leakage from these penetrations is 

72.79 cfh [cubic feet per hour]. This leakage 

rate is at a level which ensures the 30-day 

post-accident suppression pool level. 


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As mentioned above, the integrity of the 

closed system piping is verified via a 10 CFR 

50 Appendix J Type A test and is monitored 

and controlled via Technical Specification 

5.5.2. TS 5.5.2 establishes a program to 

monitor and control leakage from systems 

located outside containment that could 

contain highly radioactive fluids during a 

serious transient or accident. This program 

applies to the ECCS and RCIC systems 

affected by the proposed change and ensures 

that leakage into secondary containment via 

packing, flanges, seals, etc., is controlled. 

Leakage from these systems has been found 

to be very low, and well below the 20 gpm 

[gallons per minute] limit established for 

these systems. The proposed change is not 

expected to contribute to higher levels of 

system leakage. Normal operational 

monitoring of suppression pool level, 

operator rounds, housekeeping inspections, 

and system pressure testing further ensure 

external leakage is identified and minimized 

while suppression pool level is being 

maintained. 

A review of water leak rate test data for the 

subject CIVs showed that the valves have had 

leakage rates within the acceptance criteria. 

Testing of the valves in accordance with 

ASME [American Society of Mechanical 

Engineers] Code requirements ensure valve 

operability. 

Therefore, leakage past the CIVs is 

expected to be low and in keeping with the 

design basis for the suppression pool. 

However, the capability does exist to makeup water to the suppression pool if 

necessary. Existing Emergency Operating 

Instructions require actions if suppression 

pool level is less than the required level. 

Thus, the level of the suppression pool is 

ensured, independent of the current CIV 

water leak rate testing requirement. 

The proposed change to the scope of water 

leak rate testing for the subject valves 

maintains the existing barriers to primary 

containment leakage, and ensures that the 

suppression pool level is assured for 30 days 

during all design basis, post-accident modes 

of operation. Therefore, the plant response to 

the design basis events is unchanged, and the 

proposal 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? 

This change does not involve a significant 

reduction in a margin of safety. 

As discussed in the responses to questions 

1 and 2, the proposed change does not alter 

the plant response to existing accident 

scenarios, and does not introduce new or 

different scenarios. So the margin of safety 

from a design basis accident standpoint is 

maintained. 

Historically, the leakage rate through the 

subject valves has been determined in 

accordance with TS SR 3.6.1.3.11. This 

leakage rate has always been within the 

acceptance criteria. Quantifying leakage past 

the CIVs has been used to ensure that the 

suppression pool level is assured for 30 days 

post-accident. Under the proposed change, 

this leakage rate will not be quantified. In 

addition, closed system leakage is monitored 

and controlled by an existing Technical 

Specification program. Closed system leakage 

has been found to be very low on each of the 

units, and is currently well below the 20 gpm 

allowable. Therefore, leakage past the CIVs is 

expected to be low and in keeping with the 

design basis for the suppression pool. 

However, the capability does exist, and is 

proceduralized, to make-up water to the 

suppression pool if necessary. Thus the 

current capability to maintain adequate 

suppression pool level for 30 days postaccident is assured under the proposed 

change. 

Therefore the proposed change to the scope 

of water leak rate testing for the subject 

valves 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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: Thomas H. Boyce. 

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. 

Florida Power and Light Company, 

Docket Nos. 50–250 and 251, Turkey 

Point Plant Units 3, and 4, Miami-Dade 

County, Florida 

Date of application for amendment: 

September 1, 2009. 

Description of amendment request: 

Delay the date specified in License 

Amendments 234 and 229 for the 

implementation of the Boraflex Remedy 

in the spent fuel pools. 

Date of publication of individual 

notice in the Federal Register: 

September 15, 2009 (74 FR 47278). 

Expiration date of individual notice: 

October 15, 2009 (Public comments) and 

November 16, 2009 (Hearing requests). 

Florida Power and Light Company, 

Docket Nos. 50–250 and 50–251, Turkey 

Point Plant, Units 3 and 4, Miami-Dade 

County, Florida 

Date of application for amendments: 

July 23, 2009. 

Description of amendments request: 

Revise the scope of the inservice 

inspections required in the tubesheet 

regions of the steam generators. 

Date of publication of individual 

notice in the1 Federal Register: 

August 28, 2009 (74 FR 44405). 

Expiration date of individual notice: 

September 28, 2009 (Public comments) 

and October 27, 2009 (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 

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 


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North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Unit Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendment: 

May 28, 2009, as supplemented by letter 

dated August 3, 2009. 

Brief description of amendment: The 

amendments eliminated working hour 

restrictions from Technical 

Specification (TS) 5.2.2 for Palo Verde 

Nuclear Generating Station, Units 1, 2, 

and 3, to support compliance with the 

revisions to Title 10 of the Code of 

Federal Regulations (10 CFR), Part 26, 

‘‘Fitness for Duty Programs,’’ that 

became effective on March 31, 2008. 

The changes are consistent with the 

NRC-approved Technical Specification 

Task Force (TSTF) Standard Technical 

Specification change traveler, TSTF– 

511, Revision 0, ‘‘Eliminate Working 

Hour Restrictions from TS 5.2.2 to 

Support Compliance with 10 CFR part 

26.’’ 

Date of issuance: September 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: Unit 1–175; Unit 2– 

175; Unit 3–175. 

Facility Operating License Nos. NPF– 

41, NPF–51, and NPF–74: The 

amendments revised the Operating 

Licenses and Technical Specifications. 

Date of initial notice in Federal 

Register: July 28, 2009 (74 FR 37247). 

The supplemental letter dated August 3, 

2009, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the NRC staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 30, 

2009. 

No significant hazards consideration 

comments received: No. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendments: 

May 13, 2009. 

Brief description of amendments: The 

amendments revise Technical 

Specification (TS) 5.5.8, ‘‘Inservice 

Testing Program,’’ by incorporating TS 

Task Force Traveler (TSTF) 479, 

‘‘Changes to Reflect Revision of 10 CFR 

[Title 10 of the Code of Federal 

Regulations] 50.55a,’’ and TSTF–497, 

‘‘Limit Inservice Testing Program SR 

[Surveillance Requirement] 3.0.2 

Application to Frequencies of 2 Years or 

Less.’’ Specifically, the amendments (1) 

replace references to the American 

Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel 

Code, Section XI with the ASME Code 

for Operation and Maintenance of 

Nuclear Power Plants for inservice 

testing activities, and (2) applies the 

extension allowance of SR 3.0.2 to other 

normal and accelerated inservice testing 

frequencies of 2 years or less that were 

not included in the frequencies of the 

table listed in TS 5.5.8.a. 

Date of issuance: September 28, 2009. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment Nos.: 294 and 270. 

Renewed Facility Operating License 

Nos. DPR–53 and DPR–69: Amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 14, 2009 (74 FR 34046). 

The Commission’s related evaluation of 

these amendments is contained in a 

Safety Evaluation dated September 28, 

2009. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: 

November 20, 2008, as supplemented by 

letter dated August 12, 2009. 

Brief description of amendment: The 

amendment revised Technical 

Specification 5.6.5, ‘‘Core Operating 

Limits Report (COLR),’’ to add a 

reference to an analytical method that 

will be used to determine the core 

operating limits. The change is needed 

to support the use of GE14 fuel during 

refueling outage 15 scheduled for the 

fall of 2009. 

Date of issuance: September 29, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to Cycle 16 operation. 

Amendment No.: 166. 

Facility Operating License No. NPF– 

47: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 23, 2009 (74 FR 

4249). The supplemental letter dated 

August 12, 2009, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 29, 

2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: 

September 17, 2008, as supplemented 

by letters dated February 26, June 30, 

and September 24, 2009. 

Brief description of amendment: The 

amendment revised the Waterford 3 

Technical Specifications (TSs) to take 

credit for soluble boron in Region 1 

(cask storage pit) and Region 2 (spent 

fuel pool and refueling canal) fuel 

storage racks for the storage of both 

Standard and Next Generation Fuel 

assemblies. Two new TSs were added 

which included a surveillance that 

ensures the required boron 

concentration is maintained in the spent 

fuel storage racks and to reflect the 

results of the new criticality analysis. 

Date of issuance: September 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 223. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 14, 2009 (74 FR 17228). 

The application dated September 17, 

2008, contained an evaluation of the TS 

change in accordance with 10 CFR 

50.91(a)(1) using criteria in 10 CFR 

50.92(c), and the licensee determined 

that the change involved no significant 

hazards consideration (NSHC). 

However, based on the discussions 

between the staff and the licensee, the 

licensee provided a revised NSHC in its 

supplemental letter dated February 26, 

2009. Based on the February 26, 2009, 

revised NSHC, the staff’s proposed 


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NSHC determination was published in 

the Federal Register on April 14, 2009. 

The supplemental letters dated June 30 

and September 24, 2009, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as noticed, and did not 

change the staff’s proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 30, 

2009. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–412, 

Beaver Valley Power Station, Unit No. 2, 

Beaver County, Pennsylvania 

Date of application for amendment: 

October 10, 2008, as supplemented by 

letters dated. June 16 and July 14, 2009. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) 5.5.5 to allow an 

additional method of repair for steam 

generator (SG) tubes by installation of 

leak limiting Alloy 800 sleeves 

developed by Westinghouse and 

clarifies an existing reporting 

requirement in TS 5.6.6.2.4 concerning 

SG tube inspections. 

Date of issuance: September 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to achieving Mode 4 during 

startup from the fall 2009 refueling 

outage. 

Amendment No: 170 

Facility Operating License No. NPF– 

73. Amendment revised the License and 

TSs. 

Date of initial notice in Federal 

Register: February 17, 2009 (74 FR 

7482). The June 16 and July 14, 2009, 

supplemental letters provided clarifying 

information that was within the scope of 

the initial notice and did not change the 

initial proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 30, 

2009. 

No significant hazards consideration 

comments received: No. 

Northern States Power Company— 

Minnesota, LLC, Docket No. 50–263, 

Monticello Nuclear Generating Plant, 

Wright County, Minnesota 

Date of application for amendment: 

May 29, 2009. 

Brief description of amendment: The 

amendment changes the Technical 

Specifications, revising the applicability 

for isolation of the Reactor Water 

Cleanup System on a Standby Liquid 

Control system initiation to align with 

the modes stated in Specification 3.1.7. 

Date of issuance: September 28, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days of issuance. 

Amendment No.: 164. 

Facility Operating License No. DPR– 

22. Amendment revised the Facility 

Operating License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 28, 2009 (74 FR 37248). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated September 28, 2009. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50– 

424 and 50–425, Vogtle Electric 

Generating Plant, Units 1 and 2, Burke 

County, Georgia 

Date of application for amendments: 

May 19, 2009, as supplemented August 

28, 2009 (three submittals) and 

September 11, 2009. 

Brief description of amendments: The 

amendments revised TS 5.5.9, ‘‘Steam 

Generator (SG) Program,’’ to exclude 

portions of the tubes within the 

tubesheet from periodic SG inspections 

(establish alternate repair criteria). The 

amendments also revised TS 5.6.10, 

‘‘Steam Generator Tube Inspection 

Report,’’ to remove reference to previous 

interim alternate repair criteria and 

provide specific reporting requirements 

for Unit 1 during refueling outage (RFO) 

15 and the subsequent operating cycle, 

and for Unit 2 during RFO 14 and the 

subsequent operating cycle. 

Date of issuance: September 24, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 157 and 138. 

Renewed Facility Operating License 

Nos. NPF–68 and NPF–81: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: June 18, 2009 (74 FR 28962). 

The supplements dated August 28, 

2009, and September 11, 2009, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 24, 

2009. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

May 4, 2009. 

Brief description of amendment: The 

amendment revised the Callaway Plant 

Technical Specification (TS) 5.2.2, 

‘‘Unit Staff,’’ to eliminate working hour 

restrictions in paragraph d of TS 5.2.2 

to support compliance with Title 10 of 

the Code of Federal Regulations (10 

CFR) Part 26. The change is consistent 

with U.S. Nuclear Regulatory 

Commission (NRC)-approved Revision 0 

to TS Task Force (TSTF) Improved 

Technical Specification change traveler, 

TSTF–511, ‘‘Eliminate Working Hour 

Restrictions from TS 5.2.2 to Support 

Compliance with 10 CFR Part 26.’’ The 

availability of this TS improvement was 

announced in the Federal Register on 

December 30, 2008 (73 FR 79923), as 

part of the consolidated line item 

improvement process. 

Date of issuance: September 29, 2009. 

Effective date: As of its date of 

issuance and shall be implemented by 

October 1, 2009. 

Amendment No.: 193. 

Facility Operating License No. NPF– 

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 28, 2009 (74 FR 37250). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 29, 

2009. 

Date of initial notice in Federal 

Register: July 28, 2009 (74 FR 37250). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 29, 

2009. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 8th day 

of October 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–24915 Filed 10–19–09; 8:45 am] 

BILLING CODE 7590–01–P 


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