42926 Federal Register / Vol. 74, No. 163 / Tuesday, August 25, 2009 / Notices 

Director of the Office of Government 

Affairs, or if the incumbent is 

unavailable. 

This delegation will remain in effect 

until revoked or otherwise superseded. 

Kathleen Edwards, 

Director of Administrative Services, National 

Endowment for the Arts. 

[FR Doc. E9–20426 Filed 8–24–09; 8:45 am] 

BILLING CODE 7537–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0363] 

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 July 30, 

2009 to August 12, 2009. The last 

biweekly notice was published on 

August 11, 2009 (74 FR 40233). 

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 


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matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

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 

2007 (72 FR 49139, August 28, 2007). 

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 electronic filing Help Desk, which 

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

Eastern Time, Monday through Friday, 

excluding government holidays. The 

toll-free help line number is 1–866– 

672–7640. A person filing electronically 

may also seek assistance by sending an 

e-mail to the NRC electronic filing Help 

Desk 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 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. 

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 


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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. 

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: June 29, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

the requirements in Technical 

Specification (TS) 5.5.6, ‘‘Inservice 

Testing Program.’’ TS 5.5.6 currently 

contains references to the American 

Society of Mechanical Engineers Boiler 

and Pressure Vessel Code (ASME Code), 

Section XI as the source of requirements 

for the inservice testing (IST) of ASME 

Code Class 1, 2, and 3 pumps and 

valves. The proposed changes would 

delete the references to Section Xl of the 

ASME Code and incorporate references 

to the ASME Code for Operation and 

Maintenance of Nuclear Power Plants 

(ASME OM Code). In addition, the 

proposed amendment would address 

the applicability of Surveillance 

Requirement 3.0.2 to other normal and 

accelerated frequencies as 2 years or less 

in the IST program. These changes are 

consistent with changes identified in 

the Improved Standard Technical 

Specifications (ISTS) by Technical 

Specification Task Force Traveler 

(TSTF) Nos. 479 and 497. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the Technical 

Specification Inservice Testing Program for 

consistency with the requirements of 10 CFR 

50.55a(f)(4) for pumps and valves which are 

classified as American Society of Mechanical 

Engineers (ASME) Code Class 1, Class 2 and 

Class 3. The proposed change incorporates 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. 

The proposed changes revise TS 5.5.6 for 

RBS to conform to the requirements of 10 

CFR 50.55a(f) regarding the IST of pumps 

and valves for the third 10-Year Interval. The 

current TS reference the ASME Boiler and 

Pressure Vessel Code, Section XI, 

requirements for the IST of ASME Code Class 

1, 2, and 3 pumps and valves. The proposed 

changes would reference the ASME OM Code 

instead. This is consistent with 10 CFR 

50.55a(f). The proposed changes are 

administrative in nature. 

The proposed change does not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. They do not involve the addition or 

removal of any equipment, or any design 

changes to the facility. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises the Technical 

Specification Inservice Testing Program for 

consistency with the requirements of 10 CFR 

50.55a(f)(4) for pumps and valves which are 

classified as ASME Code Class 1, Class 2 and 

Class 3. The proposed change incorporates 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. 

The proposed TS changes do not involve 

physical changes to the facility. In addition, 

the proposed changes have no affect on plant 

configuration, or method of operation of 

plant structures, systems, or components. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change revises the Technical 

Specification Inservice Testing Program for 

consistency with the requirements of 10 CFR 

50.55a(f)(4) for pumps and valves which are 

classified as ASME Code Class 1, Class 2 and 

Class 3. The proposed change incorporates 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. 

The 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 IST of the Class 1, 2, and 3 pumps and 

valves continue to meet the appropriate 

requirements. 

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. 

Pacific Gas and Electric Company, 

Docket Nos. 50–275 and 50–323, Diablo 

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of amendment request: July 3, 

2009. 

Description of amendment request: 

The proposed amendments would 

revise the operability requirements and 

actions in Technical Specification (TS) 

3.4.15, ‘‘RCS [Reactor Coolant System] 

Leakage Detection Instrumentation,’’ 

and the associated Bases Section to 

reflect the revised TSs. 

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 reduces the time 

allowed for the plant to operate when the 

only Technical Specification (TS) 3.4.15 

operable Reactor Coolant System (RCS) 

leakage instrumentation monitor is the 

containment atmosphere gaseous 

radioactivity monitor, and revises the basis 

for operability for the containment sump 

monitors, containment atmosphere 

particulate radioactivity monitor, 

containment atmosphere gaseous 

radioactivity monitor, and the containment 

fan cooler unit condensate collection 

monitor. The proposed change increases the 

allowed operating time when all RCS leakage 

detection system instrumentation is 

inoperable. The proposed change also 

removes the word ‘‘required’’ from TS 3.4.15 

Condition A, Required Action A.2, Condition 

B, and Required Action B.2, revises TS 3.4.15 

Condition A to apply to any containment 

sump monitor, and revises the name of the 

containment fan cooler unit (CFCU) 

condensate collection monitor in the TS 

3.4.15 Actions. The monitoring of RCS 

leakage is not a precursor to any accident 

previously evaluated. The monitoring of RCS 

leakage is not used to mitigate the 

consequences of any accident previously 

evaluated. 

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 accident 

from any accident previously evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration of the plant or the 

addition of new or different type of 


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equipment. The change does not involve a 

change in how the plant is operated. 

Therefore, the proposed change does not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The change that reduces the allowed time 

of operation with only the least accurate 

containment atmosphere gaseous radiation 

monitor operable increases the margin of 

safety by increasing the likelihood that an 

increase in RCS leakage will be detected 

before it potentially results in gross failure. 

For the change that allows a limited period 

of time to restore at least one RCS leakage 

detection monitor to operable status when all 

leakage detection monitors are inoperable, 

two sources of diverse leakage detection 

capability are required to be provided during 

the limited period. Allowing a limited period 

of time to restore at least one RCS leakage 

detection instrument to operable status 

before requiring a plant shutdown avoids the 

situation of putting the plant through a 

thermal transient without RCS leakage 

monitoring. The change to TS 3.4.15 

Condition A, Required Action A.2, Condition 

B, Required Action B.2, Condition C, and 

Required Action C.2.2 is consistent with TS 

[Limiting Condition for Operation] 3.4.15 and 

does not impact the RCS leakage 

instrumentation. The revision to the TS bases 

for operability of the RCS leakage 

instrumentation monitors does not involve a 

change in the leakage instrumentation and is 

consistent with the original design of the 

leakage instrumentation. 

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 requests involve no 

significant hazards consideration. 

Attorney for licensee: Jennifer Post, 

Esq., Pacific Gas and Electric Company, 

P.O. Box 7442, San Francisco, California 

94120. 

NRC Branch Chief: Michael T. 

Markley. 

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: April 9, 

2009. 

Description of amendment request: 

The proposed amendment would 

relocate Technical Specification (TS) 

requirements pertaining to 

communications during refueling 

operations (TS 3/4.9.5), manipulator 

crane operability (TS 3/4.9.6), and crane 

travel (TS 3/4.9.7) to the Technical 

Requirements Manual. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration. The NRC staff has 

reviewed the licensee’s analysis against 

the standards of 10 CFR 50.92(c). The 

staff’s review is presented below. 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment would relocate 

TS requirements to the Technical 

Requirements Manual (TRM) which is a 

licensee-controlled document. The TS 

requirements to be relocated relate to control 

room communications during refueling, 

operability of the manipulator crane and 

auxiliary hoist for movement of control rods 

or fuel assemblies within the reactor pressure 

vessel, and control of heavy loads over fuel 

assemblies in the fuel storage pool. Once 

relocated, any future changes would be 

controlled by 10 CFR 50.59. The proposed 

amendment is administrative in nature from 

the standpoint that the current TS 

requirements would be relocated verbatim to 

the TRM. There are no physical plant 

modifications associated with this change. 

The proposed amendment would not alter 

the way any structure, system, or component 

(SSC) functions and would not alter the way 

the plant is operated. As such, the proposed 

amendment would have no impact on the 

ability of the affected SSCs to either preclude 

or mitigate an accident. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment would not 

change the design function or operation of 

the SSCs involved and would not impact the 

way the plant is operated. As such, the 

proposed change would not introduce any 

new failure mechanisms, malfunctions, or 

accident initiators not already considered in 

the design and licensing bases. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The margin of safety is associated with the 

confidence in the ability of the fission 

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

coolant pressure boundary, and containment 

structure) to limit the level of radiation to the 

public. There are no physical plant 

modifications associated with the proposed 

amendment. The proposed amendment 

would not alter the way any SSC functions 

and would not alter the way the plant is 

operated. The proposed amendment would 

not introduce any new uncertainties or 

change any existing uncertainties associated 

with any safety limit. The proposed 

amendment would have no impact on the 

structural integrity of the fuel cladding, 

reactor coolant pressure boundary, or 

containment structure. Based on the above 

considerations, the NRC staff concludes that 

the proposed amendment would not degrade 

the confidence in the ability of the fission 

product barriers to limit the level of radiation 

to the public. Therefore, the proposed change 

does not involve a significant reduction in a 

margin of safety. 

Based on this review, it appears that 

the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: February 

3, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Operating Licenses to deviate from 

certain South Texas Project Fire 

Protection Program requirements. The 

amendment will allow the performance 

of operator manual actions to achieve 

and maintain safe shutdown in the 

event of a fire in lieu of meeting circuit 

separation protection requirements of 

Title 10 of the Code of Federal 

Regulations (10 CFR), Part 50, Appendix 

R, Section III.G.2 for Fire Area 31. 

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 functions of structures, systems 

and component[s] are not impacted by the 

proposed change. The proposed change 

involves operator manual actions in response 

to a fire and will not initiate an event. The 

proposed actions do not increase the 

probability of occurrence of a fire or any 

other accident previously evaluated. 

The proposed actions are feasible and 

reliable and demonstrate that the unit can be 

safely shutdown in the event of a fire. No 

significant consequences result from the 

performance of the proposed actions. 

Therefore, the proposed change does not 

involve a significant increase in the 


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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 design functions of structures, systems 

and component[s] are not impacted by the 

proposed amendment. The proposed change 

involves operator manual actions in response 

to a fire. They do not involve new failure 

mechanisms or malfunctions that can initiate 

a new accident. 

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. 

Adequate time is available to perform the 

proposed operator manual actions to account 

for uncertainties in estimates of the time 

available and in estimates of how long it 

takes to diagnose and execute the actions. 

The actions are straightforward and do not 

create any significant concerns. The actions 

have been verified that they can be 

performed through demonstration and they 

are proceduralized. The proposed actions are 

feasible and reliable and demonstrate that the 

unit can be safely shutdown in the event of 

a fire. 

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 standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendment involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: March 3, 

2009. 

Description of amendment request: 

The proposed change would revise the 

Operating Licenses to deviate from 

certain South Texas Project Fire 

Protection Program requirements. The 

amendment will allow the performance 

of operator manual actions to achieve 

and maintain safe shutdown in the 

event of a fire in lieu of meeting circuit 

separation protection requirements of 

Title 10 of the Code of Federal 

Regulations (10 CFR), Part 50, Appendix 

R, Section III.G.2 for Fire Area 27. 

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 functions of structures, systems 

and components are not impacted by the 

proposed change. The proposed change 

involves operator manual actions in response 

to a fire, and will not initiate an event. The 

proposed actions do not increase the 

probability of occurrence of a fire or any 

other accident previously evaluated. 

The proposed actions are feasible and 

reliable and demonstrate that the unit can be 

safely shutdown in the event of a fire. No 

significant consequences result from the 

performance of the proposed actions. 

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 design functions of structures, systems 

and components are not impacted by the 

proposed amendment. The proposed change 

involves operator manual actions in response 

to a fire. They do not involve new failure 

mechanisms or malfunctions that can initiate 

a new accident. 

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 rendition in a margin of safety? 

Response: No. 

Adequate time is available to perform the 

proposed operator manual actions to account 

for uncertainties in estimates of the time 

available and in estimates of how long it 

takes to diagnose and execute the actions. 

The actions are straightforward and do not 

create any significant concerns. The actions 

have been verified that they can be 

performed through demonstration and they 

are proceduralized. The proposed actions are 

feasible and reliable and demonstrate that the 

unit can be safely shutdown in the event of 

a fare. 

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 standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendment involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

Tennessee Valley Authority, Docket No. 

50 390, Watts Bar Nuclear Plant, Unit 

1, Rhea County, Tennessee 

Date of amendment request: July 9, 

2009. 

Description of amendment request: 

The proposed amendment would allow 

use of a dedicated on-line core power 

distribution monitoring system (PDMS) 

to enhance surveillance of core thermal 

limits and would revise Technical 

Specification (TS) TS 1.1, ‘‘Definitions,’’ 

TS 3.1.8, ‘‘Rod Position Indication,’’ TS 

3.2.1, ‘‘Heat Flux Hot Channel Factor,’’ 

TS 3.2.4, ‘‘Quadrant Power Tilt Ratio 

(QPTR)’’, and TS 3.3.1, ‘‘Reactor Trip 

System (RTS) Instrumentation.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The Power Distribution Monitoring System 

(PDMS) performs essentially continuous core 

power distribution monitoring with data 

input from existing plant instrumentation. 

This system utilizes an NRC-approved 

Westinghouse proprietary computer code, i.e. 

Best Estimate Analyzer for Core Operations— 

Nuclear (BEACON), to provide data 

reduction for incore flux maps, core 

parameter analysis, load follow, operation 

simulation, and core prediction. The PDMS 

does not provide any protection or control 

system function. Fission product barriers are 

not impacted by these proposed changes. The 

proposed changes occurring with PDMS will 

not result in any additional challenges to 

plant equipment that could increase the 

probability of any previously evaluated 

accident. The changes associated with the 

PDMS do not affect plant systems such that 

their function in the control of radiological 

consequences is adversely affected. These 

proposed changes will, therefore, not affect 

the mitigation of the radiological 

consequences of any accident described in 

the Updated Final Safety Analysis Report 

(UFSAR). 

Use of the PDMS supports maintaining the 

core power distribution within required 

limits. Further, continuous on-line 

monitoring through the use of PDMS 

provides significantly more information 

about the power distributions present in the 

core than is currently available. This result 

in more time (i.e. earlier determination of an 

adverse condition developing) for operator 

action prior to having an adverse condition 

develop that could lead to an accident 

condition or to unfavorable initial conditions 

for an accident. 

Therefore, the proposed change does not 

involve a significant increase in the 


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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. 

Other than use of the PDMS to monitor 

core power distribution, implementation of 

the PDMS and associated Technical 

Specification changes has no impact on plant 

operations or safety, nor does it contribute in 

any way to the probability or consequences 

of an accident. No safety related equipment, 

safety function, or plant operation will be 

altered as a result of this proposed change. 

The possibility for a new or different type of 

accident from any accident previously 

evaluated is not created since the changes 

associated with implementation of the PDMS 

do not result in a change to the design basis 

of any plant component or system. The 

evaluation of the effects of using the PDMS 

to monitor core power distribution 

parameters shows that all design standards 

and applicable safety criteria limits are met. 

The proposed changes do not result in any 

event previously deemed incredible being 

made credible. Implementation of the PDMS 

will not result in any additional adverse 

condition and will not result in any increase 

in the challenges to safety systems. The cycle 

specific variables required by the PDMS are 

calculated using NRC approved methods. 

The Technical Specifications will continue to 

require operation within the required core 

operating limits, and appropriate actions will 

continue to be taken when or if limits are 

exceeded. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of an accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

No margin of safety is adversely affected by 

the implementation of the PDMS. The 

margins of safety provided by current 

Technical Specification requirements and 

limits remain unchanged, as the Technical 

Specifications will continue to require 

operation within the core limits that are 

based on NRC approved reload design 

methodologies. Appropriate measures exist 

to control the values of these cycle specific 

limits, and appropriate actions will continue 

to be specified and taken for when limits are 

violated. Such actions remain unchanged. 

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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: L. Raghavan. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: March 

20, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Operating License No. NPF–30 for 

Callaway Plant, Unit 1, in order to 

incorporate a change to Technical 

Specification (TS) 5.5.16, ‘‘Containment 

Leakage Rate Testing Program,’’ which 

establishes the program for leakage rate 

testing of the containment, as required 

by Title 10 of Code of Federal 

Regulations (10 CFR) Section 50.54, 

‘‘Conditions of licenses,’’ Subsection (o) 

and 10 CFR 50, Appendix J, ‘‘Primary 

Reactor Containment Leakage Testing 

for Water-Cooled Power Reactors,’’ 

Option B, ‘‘Performance Based 

Requirements,’’ as modified by 

approved exemptions. Specifically, the 

TS 5.5.16 would be revised to reflect a 

one-time 5-year deferral of the 

containment Type A integrated leak rate 

test (ILRT) from once in 10 years to once 

in 15 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. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No 

The proposed change will revise Callaway 

Plant TS 5.5.16, ‘‘Containment Leakage Rate 

Testing Program,’’ to reflect a one-time, fiveyear extension for the containment Type A 

test date to enable the implementation of a 

15-year test interval. While the containment 

is designed to contain radioactive material 

that may be released from the reactor core 

following a design basis Loss-of-Coolant 

Accident (LOCA), the test interval associated 

with Type A testing is part of ensuring the 

plant’s ability to mitigate the consequences of 

accidents described in the FSAR [Final 

Safety Analysis Report] and does not involve 

a precursor or initiator of any accident 

previously evaluated. Thus, the proposed 

change to the Type A test interval cannot 

increase the probability of an accident 

previously evaluated in the FSAR. 

Type A testing does provide assurance that 

the containment will not exceed allowable 

leakage rate criteria specified in the TS and 

will continue to perform its design function 

following an accident. However, per 

NUREG–1493, ‘‘Performance-Based 

Containment Leak-Test Program,’’ Type A 

tests identify only a few potential leakage 

paths that cannot be identified by Type B and 

C testing. The current Type B and C 

penetration test frequencies for Callaway are 

established based on performance, using the 

requirements of 10 CFR 50, Appendix J, 

Option B, and the Type B and C testing 

requirements will not be changed as a result 

of the proposed license amendment. As a 

result, with respect to the consequences of an 

accident, a risk assessment of the proposed 

change has concluded that there is an 

insignificant increase in total population 

dose rate and an insignificant increase in the 

conditional containment failure probability. 

Based on the above, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No 

The proposed change is for a one-time, 

five-year extension of the Type A test for 

Callaway Plant and will not affect the control 

parameters governing unit operation or the 

response of plant equipment to transient or 

accident conditions. The proposed change 

does not introduce new equipment, modes of 

system operation, or failure mechanisms. 

Therefore, based on the above, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No 

The Callaway Plant containment consists 

of the concrete containment building, its 

steel liner, and the penetrations through this 

structure. The structure is designed to 

contain radioactive material that may be 

released from the reactor core following a 

design basis LOCA. Additionally, this 

structure provides shielding from the fission 

products that may be present in the 

containment atmosphere following accident 

conditions. 

The containment is a prestressed, 

reinforced concrete, cylindrical structure 

with a hemispherical dome and a reinforced 

concrete base slab. The inside structure is 

lined with a carbon steel liner to ensure a 

high degree of leak tightness during operating 

and accident conditions. A post-tensioning 

system is used to prestress the cylindrical 

shell and dome. 

The concrete containment building is 

required for structural integrity of the 

containment under Design Basis Accident 

(DBA) conditions. The steel liner and its 

penetrations establish the leakage-limiting 

boundary of the containment. Maintaining 

operability of the containment will limit 

leakage of fission product radioactivity 

released from the containment to the 

environment. 

The integrity of the containment 

penetrations and isolation valves is verified 

through Type B and Type C local leak rate 

tests (LLRTs) and the overall leak tight 

integrity of the containment is verified by an 

ILRT, as required by 10 CFR 50, Appendix 

J, ‘‘Primary Reactor Containment Leakage 

Testing for Water-Cooled Power Reactors.’’ 

The existing 10-year interval at Callaway 

Plant is based on past performance. Previous 

Type A tests conducted at Callaway Plant 


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indicate that leakage from containment has 

been less than all 10 CFR 50 Appendix J, 

Option B, leakage limits. 

The proposed change for a one-time 

extension of the Type A test does not affect 

the method for Type A, B, or C testing or the 

test acceptance criteria. Type B and C testing 

will continue to be performed at the 

frequency required by Callaway Plant 

Technical Specifications. The containment 

inspections that are performed in accordance 

with the requirements of the American 

Society of Mechanical Engineers (ASME) 

Boiler and Pressure Vessel Code, Section XI, 

‘‘Inservice Inspection,’’ and 10 CFR 50.65, 

‘‘Requirements for Monitoring the 

Effectiveness of Maintenance at Nuclear 

Power Plants,’’ provide a high degree of a 

assurance that the containment will not 

degrade in a manner that is only detectable 

by Type A testing. 

In NUREG–1493, ‘‘Performance-Based 

Containment Leak-Test Program,’’ the NRC 

indicated that a 20-year extension for Type 

A testing resulted in an imperceptible 

increase in risk to the public. The NUREG– 

1493 study also concluded that, generically, 

the design containment leak rate contributes 

a very small amount to the individual risk 

and that the decrease in Type A testing 

frequency would have a minimal affect on 

this risk. AmerenUE has conducted risk 

assessments to determine the impact of a 

one-time change to the Callaway Plant Type 

A test schedule from a baseline value of once 

in 10 years to once in 15 years for the risk 

measures of Large Early Release Frequency 

(LERF), Total Population Dose, and 

Conditional Containment Failure Probability 

(CCFP). The results of the risk assessments 

indicate that the proposed change to the 

Callaway Plant Type A test schedule has a 

minimal impact on public risk. 

Based on the above and on previous Type 

A test results for the Callaway Plant 

containment, the current containment 

surveillance program, and the results of the 

AmerenUE risk assessment, there is no 

reduction in the effectiveness of the Callaway 

Plant containment as a barrier to the release 

of the post-accident containment atmosphere 

to the public or to personnel in the Control 

Room. Thus, 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: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: May 4, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.7.3, 

‘‘Main Feedwater Isolation Valves 

(MFIVs) and Main Feedwater Regulating 

Valves (MFRVs), and Main Feedwater 

Regulating Valve Bypass Valves 

(MFRVBVs),’’ so that the limiting 

condition for operation (LCO) and 

Applicability more accurately reflect the 

conditions for when the LCO should be 

applicable and more effectively provide 

appropriate exceptions to the 

Applicability for certain valve 

configurations. The amendment would 

incorporate other minor changes; the 

title to TS 3.7.3 and the header for each 

TS page would be revised, and the 

exception footnotes in TS Table 3.3.2– 

1 of TS 3.3.2, ‘‘ESFAS [Engineered 

Safety Features Actuation System] 

Instrumentation,’’ would be revised to 

improve the application of existing 

notes and/or incorporate more 

appropriate notes as applicable. 

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 changes do not alter any 

design or operating limits, nor do they 

physically alter safety-related systems, nor do 

they affect the way in which safety-related 

systems perform their functions. The 

proposed changes do not change accident 

initiators or precursors assumed or 

postulated in the FSAR [Final Safety 

Analysis Report]-described accident 

analyses, nor do they alter the design 

assumptions, conditions, and configuration 

of the facility or the manner in which the 

plant is normally operated and maintained. 

The proposed changes do not alter or prevent 

the ability of structures, systems, and 

components (SSCs) from performing their 

intended functions to mitigate the 

consequences of an initiating event within 

the assumed acceptance limits. With specific 

regard to the proposed TS changes, although 

the changes involve the exceptions contained 

in the Applicability of TS 3.7.3 as well as the 

notes attached to TS Table 3.3.2–1 (which are 

themselves exceptions), the provisions of the 

exceptions and notes would continue to be 

based on the premise that adequate isolation 

or isolation capability exists for the main 

feedwater lines, i.e., that the required safety 

function is performed or capable of being 

performed as required or assumed for 

mitigation of the applicable postulated 

accidents. 

All accident analysis acceptance criteria 

will therefore continue to be met with the 

proposed changes. The proposed changes 

will not affect the source term, containment 

isolation, or radiological release assumptions 

used in evaluating the radiological 

consequences of an accident previously 

evaluated. The proposed changes will not 

alter any assumptions or change any 

mitigation actions in the radiological 

consequence evaluations in the FSAR. The 

applicable radiological dose acceptance 

criteria will continue to be met. Overall 

protection system performance will remain 

within the bounds of the previously 

performed accident analyses. 

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 

There are no proposed design changes, nor 

are there any changes in the method by 

which any safety-related plant structure, 

system, or component (SSC) performs its 

specified safety function. The proposed 

changes will not affect the normal method of 

plant operation or change any operating 

parameters. No equipment performance 

requirements will be affected. The proposed 

changes will not alter any assumptions made 

in the safety analyses. No new accident 

scenarios, transient precursors, failure 

mechanisms, or limiting single failures will 

be introduced as a result of this amendment. 

There will be no adverse effect or challenges 

imposed on any safety-related system as a 

result of this amendment. The proposed 

amendment will not alter the design or 

performance of the 7300 Process Protection 

System, Nuclear Instrumentation System, or 

Solid State Protection System used in the 

plant protection systems. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No 

There will be no effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, heat flux hot channel factor 

(FQ), nuclear enthalpy rise hot channel factor 

(FDH), loss of coolant accident peak cladding 

temperature (LOCA PCT), peak local power 

density, or any other margin of safety. The 

applicable radiological dose consequence 

acceptance criteria for design-basis transients 

and accidents will continue to be met. The 

proposed changes do not eliminate any 

surveillances or alter the frequency of 

surveillances required by the Technical 

Specifications. None of the acceptance 

criteria for any accident analysis will be 

changed. 


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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: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: May 4, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.7.2, 

‘‘Main Steam Isolation Valves (MSIVs),’’ 

to add the main steam isolation valve 

bypass valves (MSIVBVs) and main 

steam low point drain isolation valves 

(MSLPDIVs) to the scope of the TS. In 

addition, the proposed amendment 

would make editorial changes to the 

title and header on each page of TS 

3.7.2, and would incorporate other 

minor changes to revise exception 

footnote (i) in TS Table 3.3.2–1 of TS 

3.3.2, ‘‘ESFAS [Engineered Safety 

Features Actuation System] 

Instrumentation,’’ to remove the MSIVs 

from the footnote such that the footnote 

only addresses the MSIVBVs and 

MSLPDIVs. The MSIVs would be 

addressed in new exception footnote (k) 

added to TS Table 3.3.2–1. 

The proposed amendment would add 

new TS 3.7.19, ‘‘Secondary System 

Isolation Valves (SSIVs),’’ which would 

provide limiting conditions for 

operation (LCOs) and surveillance 

requirements for the SSIVs, steam 

generator chemical injection isolation 

valves (SGCIIVs), steam generator 

blowdown isolation valves (SGBSIVs), 

and steam generator sample line 

isolation valves (SGBSSIVs). New 

Function 10, ‘‘Steam Generator 

Blowdown System and Sample Line 

Isolation Valve Actuation,’’ would be 

added to TS Table 3.3.2–1. The 

SGBSIVs and SGBSSIVs would be 

addressed in new exception footnote (t) 

added to Table 3.3.2–1 for Function 10. 

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 adds requirements to 

the TS to ensure that systems and 

components are maintained consistent with 

the safety analysis and licensing basis. 

Requirements are incorporated into the TS 

for secondary system isolation valves. These 

changes do not involve any design or 

physical changes to the facility, including the 

SSIVs themselves. The design and functional 

performance requirements, operational 

characteristics, and reliability of the SSIVs 

are unchanged. There is no impact on the 

design safety function of MSIVs, MSIVBVs, 

MSLPDIVs, MFIVs [main feedwater isolation 

valves], MFRVs [main feedwater regulating 

valves] or MFRVBVs [MFRV bypass valves] 

to close (either as an accident mitigator or as 

a potential transient initiator). Since no 

failure mode or initiating condition that 

could cause an accident (including any plant 

transient) evaluated per the FSAR [Final 

Safety Analysis Report]-described safety 

analyses is created or affected, the change 

cannot involve a significant increase in the 

probability of an accident previously 

evaluated. 

With regard to the consequences of an 

accident and the equipment required for 

mitigation of the accident, the proposed 

changes involve no design or physical 

changes to components in the main steam 

supply system or feedwater system. There is 

no impact on the design safety function of 

MSIVs, MSIVBVs, MSLPDIVs, MFIVs, 

MFRVs, or MFRVBVs or any other equipment 

required for accident mitigation. Adequate 

equipment availability would continue to be 

required by the TS. The consequences of 

applicable, analyzed accidents (such as a 

main steam line break [or] feedline break) are 

not impacted by the proposed changes. 

The changes to TS 3.3.2, TS Table 3.3.2– 

1, and exception footnotes associated with 

Table Function 4 and New Function 10 

maintain consistency with the Applicability 

of revised TS 3.7.2 and new TS 3.7.19. 

Maintaining TS 3.3.2 and TS Table 3.3.2–1 

consistent with the Applicability of TS 3.7.2 

and TS 3.7.19 is consistent with the 

Westinghouse Standard Technical 

Specifications. 

These changes involve no physical changes 

to the facility and do not adversely affect the 

availability of the safety functions assumed 

for the MSIVs, MSIVBVs, MSLPDIVs, and 

SSIVs. Therefore, they do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Based on the above considerations, the 

proposed changes do 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 add requirements to 

the TS that support or ensure the availability 

of the safety functions assumed or required 

for the MSIVs, MSIVBVs, MSLPDIVs, and 

SSIVs. The changes do not involve a physical 

alteration of the plant (no new or different 

type of equipment will be installed) or 

changes in controlling parameters. 

Additional requirements are being imposed, 

but they are consistent with the assumptions 

made in the safety analysis and licensing 

basis. The addition of Conditions, Required 

Actions and Completion Times to TS for the 

MSIVBVs, MSLPDIVs, and SSIVs does not 

involve a change in the design, configuration, 

or operational characteristics of the plant. 

Further, the proposed changes do not involve 

any changes in plant procedures for ensuring 

that the plant is operated within analyzed 

limits. As such, no new failure modes or 

mechanisms that could cause a new or 

different kind of accident from any 

previously evaluated are introduced. 

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? 

Response: No. 

The proposed addition of Conditions, 

Required Actions and Completion Times for 

SSIVs, MSIVBVs, and MSLPDIVs, as well as 

the proposed change to the LCO and 

Applicability for TS 3.7.2 and the proposed 

new TS 3.7.19 (and the corresponding 

changes to TS 3.3.2, ‘‘ESFAS 

Instrumentation’’) does not alter the manner 

in which safety limits or limiting safety 

system settings are determined. No changes 

to instrument/system actuation setpoints are 

involved. The safety analysis acceptance 

criteria are not impacted and the proposed 

change will not permit plant operation in a 

configuration outside the design basis. The 

changes are consistent with the safety 

analysis and licensing basis for the facility. 

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: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: June 1, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

the Limiting Condition for Operation 

(LCO) Applicability Note for Technical 

Specification (TS) 3.3.9, ‘‘Boron 

Dilution Mitigation System (BDMS).’’ 


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The LCO Applicability Note would be 

revised to more explicitly define what 

the term ‘‘during reactor startup’’ means 

in MODES 2 and 3. This revision to the 

Applicability Note is proposed to clarify 

the situations during which the BDMS 

signal may be blocked. 

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. 

Overall protection system performance will 

remain within the bounds of the previously 

performed accident analyses since there are 

no design changes. All design, material, and 

construction standards that were applicable 

prior to this amendment request will be 

maintained. There will be no changes to any 

design or operating limits. 

The proposed change will not adversely 

affect accident initiators or precursors [or] 

adversely alter the design assumptions, 

conditions, and configuration of the facility 

or the manner in which the plant is operated 

and maintained. There are no design or 

operating changes to the reactor makeup 

water system (RMWS), the reactor makeup 

control system (RMCS), or the chemical and 

volume control system (CVCS). There will be 

no decrease in the boron concentration of the 

boric acid tanks. There will be no changes to 

the BDMS setpoint or the operation of the 

BDMS, other than the limited durations 

during which flux multiplication signal 

blocking would be allowed. Therefore, there 

will be no changes that would serve to 

increase the likelihood of occurrence of an 

inadvertent boron dilution event. 

The proposed change will not alter or 

prevent the ability of structures, systems, and 

components (SSCs) from performing their 

intended functions to mitigate the 

consequences of an initiating event within 

the applicable acceptance limits. Exceptions 

to Technical Specification requirements are 

allowed and, in fact, rather commonplace 

when plant operation would otherwise be 

restricted in a manner that is not 

commensurate with the desired safety 

objective, especially when those exceptions 

are of short duration and are accompanied by 

compensatory measures. 

The proposed change does not physically 

alter safety-related systems [or] affect the way 

in which safety-related systems perform their 

functions. 

The inadvertent boron dilution analysis 

acceptance criteria will continue to be met 

with the proposed change, with 

consideration given to the fact that the 

current licensing basis analyses do not 

assume concurrent rod withdrawal in the 

MODES 2 and 3 boron dilution analyses. The 

licensing basis analyses assume that positive 

reactivity insertion is being added by a single 

method, i.e., boron dilution. The MODE 2 

licensing basis analysis of an inadvertent 

boron dilution event in FSAR [Final Safety 

Analysis Report] Section 15.4.6 assumes that 

the shutdown banks are fully withdrawn and 

that the control banks are withdrawn to the 

0% power rod insertion limits depicted in 

the COLR [Core Operating Limits Report]. 

The MODE 2 analysis credits operator action 

to swap the charging suction source after an 

automatic reactor trip, and corresponding rod 

insertion, on high source range neutron flux. 

The MODE 3 licensing basis analysis credits 

automatic mitigation by the BDMS with 

steady state initial conditions and static 

initial rod positions (all shutdown and 

control banks are fully inserted other than 

the single most reactive rod which is 

assumed to be fully withdrawn) at bounding 

RCS [reactor coolant system] T–avg values at 

either end of MODE 3. Neither the analysis 

nor the BDMS design basis assumes that the 

system protects against a rod withdrawal 

event. 

The proposed change will not affect the 

source term, containment isolation, or 

radiological release assumptions used in 

evaluating the radiological consequences of 

an accident previously evaluated. The 

applicable radiological dose criteria will 

continue to be met. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

There are [neither] proposed design 

changes nor are there any changes in the 

method by which any safety-related plant 

structure, system, or component (SSC) 

performs its specified safety function. The 

proposed change will not affect the normal 

method of plant operation or change any 

operating parameters. Equipment 

performance necessary to fulfill safety 

analysis missions will be unaffected. The 

proposed change will not alter any 

assumptions required to meet the safety 

analysis acceptance criteria. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures will be introduced as a result 

of this amendment. There will be no adverse 

effect or challenges imposed on any safetyrelated system as a result of this amendment. 

The proposed amendment will not alter the 

design or performance of the 7300 Process 

Protection System, Nuclear Instrumentation 

System, or Solid State Protection System 

used in the plant protection systems. 

The proposed change does not, therefore, 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

There will be no effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, heat flux hot channel factor 

(FQ), nuclear enthalpy rise hot channel factor 

(FDH), loss of coolant accident peak cladding 

temperature (LOCA PCT), peak local power 

density, or any other margin of safety. Modespecific required shutdown margins in the 

COLR will not be changed. The applicable 

radiological dose consequence acceptance 

criteria will continue to be met. 

The proposed change does not eliminate 

any surveillances or alter the frequency of 

surveillances required by the Technical 

Specifications. None of the acceptance 

criteria for any accident analysis will be 

changed. 

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: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: July 10, 

2009. 

Description of amendment request: 

The proposed amendment would delete 

the Technical Specification (TS) 

requirements for the containment 

hydrogen recombiners and hydrogen 

monitors. The proposed TS changes 

support implementation of the revision 

to Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.44, 

‘‘Standards for Combustible Gas Control 

System in Light-Water-Cooled Power 

Reactors,’’ that became effective on 

October 16, 2003. The proposed changes 

are consistent with Revision 1 of the 

NRC-approved Industry/Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–447, 

‘‘Elimination of Hydrogen Recombiners 

and Change to Hydrogen and Oxygen 

Monitors.’’ 

The NRC staff issued a notice of 

opportunity for public comments on 

TSTF–447, Revision 1, published in the 

Federal Register on August 2, 2002 (67 

FR 50374), soliciting comments on a 

model safety evaluation (SE) and a 

model no significant hazards 

consideration (NSHC) determination for 

the elimination of requirements for 

hydrogen recombiners, and hydrogen 

and oxygen monitors from TS. Based on 

its evaluation of the public comments 


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received, the NRC staff made 

appropriate changes to the models and 

included final versions in a notice of 

availability published in the Federal 

Register on September 25, 2003 (68 FR 

55416), regarding the adoption of TSTF– 

447, Revision 1, as part of the NRC’s 

consolidated line item improvement 

process (CLIIP). 

In addition to the changes related to 

requirements for the hydrogen 

recombiners and monitors, this 

amendment application includes four 

unrelated, minor changes to correct 

typographical errors identified in 

Callaway’s TS. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC adopted 

by the licensee is presented below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The revised 10 CFR 50.44 no longer defines 

a design-basis loss-of-coolant accident 

(LOCA) hydrogen release, and eliminates 

requirements for hydrogen control systems to 

mitigate such a release. The installation of 

hydrogen recombiners and/or vent and purge 

systems required by 10 CFR 50.44(b)(3) was 

intended to address the limited quantity and 

rate of hydrogen generation that was 

postulated from a design-basis LOCA. The 

Commission has found that this hydrogen 

release is not risk-significant because the 

design-basis LOCA hydrogen release does not 

contribute to the conditional probability of a 

large release up to approximately 24 hours 

after the onset of core damage. In addition, 

these systems were ineffective at mitigating 

hydrogen releases from risk-significant 

accident sequences that could threaten 

containment integrity. 

With the elimination of the design-basis 

LOCA hydrogen release, hydrogen monitors 

are no longer required to mitigate designbasis accidents and, therefore, the hydrogen 

monitors do not meet the definition of a 

safety-related component as defined in 10 

CFR 50.2. RG [Regulatory Guide] 1.97 

Category 1 is intended for key variables that 

most directly indicate the accomplishment of 

a safety function for design-basis accident 

events. The hydrogen monitors no longer 

meet the definition of Category 1 in RG 1.97. 

As part of the rulemaking to revise 10 CFR 

50.44 the Commission found that Category 3, 

as defined in RG 1.97, is an appropriate 

categorization for the hydrogen monitors 

because the monitors are required to 

diagnose the course of beyond design-basis 

accidents. 

The regulatory requirements for the 

hydrogen monitors can be relaxed without 

degrading the plant emergency response. The 

emergency response, in this sense, refers to 

the methodologies used in ascertaining the 

condition of the reactor core, mitigating the 

consequences of an accident, assessing and 

projecting offsite releases of radioactivity, 

and establishing protective action 

recommendations to be communicated to 

offsite authorities. Classification of the 

hydrogen monitors as Category 3 and 

removal of the hydrogen monitors from TS 

will not prevent an accident management 

strategy through the use of the SAMGs 

[severe accident management guidelines], the 

emergency plan (EP), the emergency 

operating procedures (EOP), and site survey 

monitoring that support modification of 

emergency plan protective action 

recommendations (PARs). 

Therefore, the elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, does not involve a significant 

increase in the probability or the 

consequences of any accident previously 

evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Previously 

Evaluated 

The elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen monitor equipment are not 

considered accident precursors, nor does 

their existence or elimination have any 

adverse impact on the pre-accident state of 

the reactor core or post accident confinement 

of radionuclides within the containment 

building. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, in light of existing plant equipment, 

instrumentation, procedures, and programs 

that provide effective mitigation of and 

recovery from reactor accidents, results in a 

neutral impact to the margin of safety. 

The installation of hydrogen recombiners 

and/or vent and purge systems required by 

10 CFR 50.44(b)(3) was intended to address 

the limited quantity and rate of hydrogen 

generation that was postulated from a designbasis LOCA. The Commission has found that 

this hydrogen release is not risk-significant 

because the design-basis LOCA hydrogen 

release does not contribute to the conditional 

probability of a large release up to 

approximately 24 hours after the onset of 

core damage. 

Category 3 hydrogen monitors are adequate 

to provide rapid assessment of current 

reactor core conditions and the direction of 

degradation while effectively responding to 

the event in order to mitigate the 

consequences of the accident. The intent of 

the requirements established as a result of the 

[Three Mile Island], Unit 2 accident, can be 

adequately met without reliance on safetyrelated hydrogen monitors. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TS 

will not result in a significant reduction in 

their functionality, reliability, and 

availability. 

The NRC staff has reviewed the 

analysis adopted by the licensee 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: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Virginia Electric and Power Company, 

Docket No. 50–338 North Anna Power 

Station, Unit No. 1, Louisa County, 

Virginia 

Date of amendment request: July 23, 

2009 

Description of amendment request: 

The proposed change, a one-time 

extension to the Completion Time (CT) 

of Technical Specification 3.8.9 

Condition A, will provide an 

opportunity to fully investigate the 

extent of the damaged breaker and its 

condition to ensure continued bus 

reliability for the remainder of the 

operating cycle. 

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 license amendment 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

The proposed change does not alter any 

plant equipment or operating practices in 

such a manner that the probability of an 

accident is significantly increased. The 

proposed change will not alter assumptions 

relative to the mitigation of an accident or 

transient event. Manual operator actions in 

the event of an SGTR have been identified 

during the one-time extended CT for the 1J1 

[Motor Control Center] MCC outage. A riskinformed evaluation of these operator actions 

has been performed and the increase in 

annual Core Damage and Large Early Release 

Frequencies associated with the proposed 

change in the Technical Specification CT are 

characterized as ‘‘small changes’’ by 

Regulatory Guide (RG) 1.174. The 

Incremental Conditional Core Damage and 

Large Early Release Probabilities [ICCDP and 

ICLERP] associated with the proposed 


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Technical Specification CT meet the 

acceptance criteria in Regulatory Guide 

1.177. 

The ICCDP and ICLERP are 1.01 E–7 per 

year and 9.86E–9 per year, respectively. 

These results are below the RG 1.177 limits 

of 5E–7 for ICCDP and 5E–8 for ICLERP. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed license amendment 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated? 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. 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? 

The systems’ design and operation are not 

affected by the proposed change. The safety 

analysis acceptance criteria stated in the 

Updated Final Safety Analysis Report is not 

impacted by the change. Redundancy and 

diversity of the electrical distribution system 

will be maintained with the exception of the 

MCCs 1J 1–2N and 2S. The proposed change 

will not allow plant operation in a 

configuration outside the design basis. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

The Nuclear Regulatory Commission 

(NRC) staff has reviewed the licensee’s 

analysis and, based on this review, it 

appears that the three standards of 

50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., Millstone 

Power Station, Building 475, 5th Floor, 

Rope Ferry Road, Rt. 156, Waterford, 

Connecticut 06385 

NRC Branch Chief: Undine Shoop. 

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. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–369 and 50–370, McGuire 

Nuclear Station, Units 1 and 2, 

Mecklenburg County, North Carolina 

Date of amendment request: March 

20, 2008, as supplemented by letters 

dated May 28, 2008, October 6, 2008, 

December 17, 2008, and February 12, 

2009. 

Brief description of amendment 

request: The proposed amendments 

would revise the McGuire licensing 

basis by adopting the Alternative Source 

Term (AST) radiological analysis 

methodology as allowed by 10 CFR 

50.67, Accident Source Term, for the 

Loss of Coolant Accident. This 

amendment request represents full 

scope implementation of the AST as 

described in Nuclear Regulatory 

Commission (NRC) Regulatory Guide 

1.183, ‘‘Alternative Radiological Source 

Terms for Evaluating Design Basis 

Accidents at Nuclear Power Reactors, 

Revision 0.’’ 

Date of publication of individual 

notice in Federal Register: February 27, 

2009 (74 FR 9009). 

Expiration date of individual notice: 

April 28, 2009. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 

2, York County, South Carolina 

Date of amendment request: June 23, 

2008. 

Brief description of amendment 

request: The amendments revise the 

Technical Specifications (TSs) for 

Catawba Nuclear Station, Units 1 and 2. 

This request modifies the subject TS 

and Bases by changing the logic 

configuration of TS Table 3.3.2–1, 

‘‘Engineered Safety Feature Actuation 

System Instrumentation’’, Function 5.b. 

(5), ‘‘Turbine Trip and Feedwater 

Isolation, Feedwater Isolation, Doghouse 

Water Level—High High.’’ The existing 

one-out-of-one (1⁄1) logic per train per 

doghouse is being modified to a twoout-of-three (2⁄3) logic per train per 

doghouse. The proposed change will 

improve the overall reliability of this 

function and will reduce the potential 

for spurious actuations. 

Date of publication of individual 

notice in Federal Register: February 24, 

2009 (74 FR 8276). 

Expiration date of individual notice: 

April 27, 2009. 

Duke Energy Carolinas, LLC, et al., 

Docket No. 50–414, Catawba Nuclear 

Station, Unit 2, York County, South 

Carolina 

Date of amendment request: 

November 13, 2008. 

Brief description of amendment 

request: The amendment proposes a 

one-cycle revision to the Technical 

Specifications to incorporate an interim 

alternate repair criterion for steam 

generator tube repair criteria during the 

End of Cycle 16 refueling outage and 

subsequent cycle 17 operation. 

Date of publication of individual 

notice in Federal Register: February 

24, 2009 (74 FR 8278). 

Expiration date of individual notice: 

April 27, 2009. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Steam Electric Station, 

Units 1 and 2, Somervell County, Texas 

Date of amendment request: June 8, 

2009. 

Brief description of amendment 

request: The proposed amendment 

would revise Technical Specification 

(TS) 5.5.9.2, ‘‘Unit 1 Model D76 and 

Unit 2 Model D5 Steam Generator (SG) 

Program,’’ to exclude portions of the 

CPSES, Unit 2 Model D5 SG below the 

top of the SG tubesheet from periodic 

SG tube inspections. In addition, the 

proposed amendment would revise TS 

5.6.9, ‘‘Unit 1 Model D76 and Unit 2 

Model D5 Steam Generator Tube 

Inspection Report,’’ to include reporting 

requirements specific to the permanent 

alternate repair criteria for CPSES, Unit 

2. The amendment request is supported 

by Westinghouse WCAP–17072–P, ‘‘H*: 

Alternate Repair Criteria for the Tube 

Sheet Expansion Region in Steam 

Generators with Hydraulically 

Expanded Tubes (Model D5),’’ May 

2009. 

Date of publication of individual 

notice in Federal Register: July 23, 

2009 (74 FR 36533). 

Expiration date of individual notice: 

September 21, 2009. 

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 


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Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (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@nrc.gov. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendments: 

April 23, 2009. 

Brief description of amendments: The 

amendments revise the Technical 

Specifications (TSs) by removing 

working hour restrictions from TS 5.2.2 

to support compliance with recent 

revisions to Title 10 of the Code of 

Federal Regulations, Part 26, Subpart I. 

The amendments are consistent with the 

guidance contained in Nuclear 

Regulatory Commission (NRC) approved 

Technical Specifications Task Force 

Traveler 511 (TSTF–511). This TS 

improvement was made available by the 

NRC on December 30, 2008 (73 FR 

79923) as part of the consolidated line 

item improvement process. 

Date of issuance: August 6, 2009. 

Effective date: As of the date of 

issuance to be implemented with the 

implementation of the new 10 CFR Part 

26, Subpart I requirements. 

Amendment Nos.: 292 and 268. 

Renewed Facility Operating License 

Nos. DPR–53 and DPR–69: Amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 2, 2009 (74 FR 26430). 

The Commission’s related evaluation 

of these amendments is contained in a 

Safety Evaluation dated August 6, 2009. 

No significant hazards consideration 

comments received: No. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 

2, York County, South Carolina 

Date of application for amendments: 

July 14, 2008. 

Brief description of amendments: The 

changes revised Technical 

Specifications (TSs) Section 3.7.10, 

‘‘Control Room Area Ventilation,’’ its 

associated Bases, and TS Section 5.5 

‘‘Programs and Manuals.’’ This LAR 

institutes the Control Room Habitability 

Program. 

The changes are consistent with NRCapproved Industry Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–448, Revision 3, 

‘‘Control Room Habitability Program.’’ 

The availability of this TS improvement 

was announced in the Federal Register 

on January 17, 2007, as part of the 

Consolidated Line-Item Improvement 

Process (CLIIP). The amendments also 

authorized a change to the Catawba 

Updated Final Safety Analysis Report 

(UFSAR). 

Date of issuance: July 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 250 and 245. 

Facility Operating License Nos. NPF– 

35 and NPF–52: Amendments revised 

the licenses and the technical 

specifications.XXX 

Date of initial notice in Federal 

Register: June 2, 2009 (74 FR 26431). 

The Commission’s related evaluation, 

State consultation, and final no 

significant hazards consideration 

determination of the amendments is 

contained in a Safety Evaluation dated 

July 30, 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: January 

21, 2009, as supplemented by letters 

dated January 23 and June 22, 2009. 

Brief description of amendment: The 

amendment modified the Technical 

Specifications (TSs) to adopt U.S. 

Nuclear Regulatory Commission (NRC)- 

approved TS Task Force (TSTF) change 

travelers TSTF–163, TSTF–222, TSTF– 

230, and TSTF–306, and made two 

minor administrative corrections. 

Date of issuance: August 11, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 165. 

Facility Operating License No. NPF– 

47: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 24, 2009 (74 FR 

12392). The supplemental letters dated 

January 23 and June 22, 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 August 11, 

2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 

50–313, Arkansas Nuclear One, Unit 

No. 1 (ANO1), Pope County, Arkansas 

Entergy Operations, Inc., Docket No. 

50–368, Arkansas Nuclear One, Unit 

No. 2 (ANO2), Pope County, Arkansas 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. 

FitzPatrick Nuclear Power Plant (JAF), 

Oswego County, New York 

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 (GGNS), Claiborne 

County, Mississippi 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–247 and 50–286, Indian 

Point Nuclear Generating Unit Nos. 2 

and 3 (IP2 and IP3), Westchester 

County, New York 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Plant 

(PAL), Van Buren County, Michigan 


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42938 Federal Register / Vol. 74, No. 163 / Tuesday, August 25, 2009 / Notices 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station (PIL), Plymouth County, 

Massachusetts 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 

50–458, River Bend Station, Unit 1 

(RBS), West Feliciana Parish, Louisiana 

Entergy Operations, Inc., Docket No. 

50–382, Waterford Steam Electric 

Station, Unit 3 (W3), St. Charles Parish, 

Louisiana 

Date of application for amendment: 

April 27, 2009, as supplemented July 

10, 2009. 

Brief description of amendment: The 

amendments deleted those portions of 

the Technical Specifications (TSs) 

superseded by Title 10 of the Code of 

Federal Regulations (10 CFR) Part 26, 

Subpart I, consistent with U.S. Nuclear 

Regulatory Commission (NRC)-approved 

TS Task Force (TSTF) 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: August 4, 2009. 

Effective date: As of the date of 

issuance and shall be implemented by 

October 1, 2009. 

Amendment Nos.: ANO1—237; 

ANO2—285; JAF—295; GGNS—183; 

IP2—261; IP3—240; PAL—238; PIL— 

233; RBS—164; and W3—221. 

Facility Operating License Nos. DPR– 

51 (ANO1), NPF–6 (ANO2), DPR–59 

(JAF), NPF–29 (GGNS), DPR–26 (IP2), 

DPR–64 (IP3), DPR–20 (PAL), DPR–35 

(PIL), NPF–47 (RBS), and NPF–38 (W3): 

The amendments revised the Facility 

Operating Licenses and Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 2, 2009 (74 FR 26432). 

The supplement dated July 10, 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 amendments is contained in a 

Safety Evaluation dated August 4, 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 January 8, March 18, 

and June 30, 2009. 

Brief description of amendment: The 

amendment revised the Operating 

License and modified Technical 

Specification (TS) 3⁄4.3.1 and Note 2 of 

TS Table 4.3–1. The changes result in 

the addition of conservatism to Core 

Protection Calculator power indications 

when calibrations are required in 

certain conditions. 

Date of issuance: August 10, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 222. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

65695). The supplemental letters dated 

January 8, March 18, and June 30, 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 August 10, 

2009. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, 

LaSalle County Station, Units 1 and 2, 

LaSalle County, Illinois 

Exelon Generation Company, LLC, and 

PSEG Nuclear LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of application for amendments: 

June 9, 2008, as supplemented by letter 

dated March 30, 2009. 

Brief description of amendments: The 

amendments revise the Technical 

Specification (TS) surveillance 

requirement (SR) frequency in TS 3.1.3, 

‘‘Control Rod OPERABILITY.’’ The 

amendments also clarify the 

requirement to fully insert all insertable 

control rods for the limiting condition 

for operation in TS 3.3.1.2, Required 

Action E.2, ‘‘Source Range Monitoring 

Instrumentation’’ (Clinton Power 

Station only). Finally, the amendments 

revise Example 1.4–3 in Section 1.4, 

‘‘Frequency,’’ to clarify the applicability 

of the 1.25 surveillance test interval 

extension. 

Date of issuance: August 11, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 188, 232/225, 193/ 

180, 272/276, 244/239. 

Facility Operating License Nos. NPF– 

62, DPR–19, DPR–25, NPF–11, NPF–18, 

DPR–44, DPR–56, DPR–29, DPR–30: The 

amendments revised the Technical 

Specifications/Licenses. 

Date of initial notice in Federal 

Register: August 12, 2009 (73 FR 

46928) The March 30, 2009, supplement 

contained clarifying information and 

did not change the NRC staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 11, 

2009. 

No significant hazards consideration 

comments received: No. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Steam Electric Station, 

Unit Nos. 1 and 2 (CPSES), Somervell 

County, Texas 

Date of amendment request: April 1, 

2009, as supplemented by letter dated 

July 9, 2009. 

Brief description of amendments: The 

amendments deleted Technical 

Specification (TS) 5.2.2.d, in TS 5.2.2, 

‘‘Unit Staff,’’ regarding the requirement 

to develop and implement 

administrative procedures to limit the 

working hours of personnel who 

perform safety-related functions. In 

addition, paragraphs e and f of TS 5.2.2 

were renumbered to d and e and in TS 

5.2.2.b the reference to 5.2.2.f was 

revised to 5.2.2.e to reflect the removal 

of paragraph d of TS 5.2.2. 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: August 7, 2009. 


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Federal Register / Vol. 74, No. 163 / Tuesday, August 25, 2009 / Notices 42939 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: Unit 1–148; Unit 

2–148. 

Facility Operating License Nos. NPF– 

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: May 19, 2009 (74 FR 23445). 

The supplemental letter dated July 9, 

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 amendments is contained in a 

Safety Evaluation dated August 7, 2009. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket Nos. 50–220 and 50–410, Nine 

Mile Point Nuclear Station, Unit Nos. 1 

and 2 (NMP 1 and 2), Oswego County, 

New York 

Date of application for amendment: 

February 11, 2009. 

Brief description of amendments: The 

amendments delete those portions of the 

Technical Specifications (TSs) 

superseded by Title 10 of the Code of 

Federal Regulations (10 CFR), Part 26, 

Subpart I. This change is consistent 

with Nuclear Regulatory Commission 

(NRC) approved Technical Specification 

Task Force (TSTF) Improved Standard 

Technical Specifications Change 

Traveler TSTF–511, Revision 0, 

‘‘Eliminate Working Hour Restrictions 

from TS 5.2.2 to Support Compliance 

with 10 CFR Part 26.’’ These changes 

were described in a Notice of 

Availability for Consolidated Line Item 

Improvement Process TSTF–511 

published in the Federal Register on 

December 30, 2008 (73 FR 79923). 

Date of issuance: July 27, 2009. 

Effective date: As of the date of 

issuance to be implemented by October 

1, 2009. 

Amendment Nos.: 203 and 131. 

Renewed Facility Operating License 

Nos. DPR–063 and NPF–069: The 

amendments revise the License and TSs. 

Date of initial notice in Federal 

Register: April 21, 2009 (73 FR 18255). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 27, 2009. 

No significant hazards consideration 

comments received: No. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: 

September 2, 2008. 

Brief description of amendments: The 

amendment approved the licensee’s 

request to incorporate a revision in the 

Updated Final Safety Analysis Report 

(UFSAR) Section 13.7.2.3, ‘‘PRA Risk 

Categorization,’’ to add a separate set of 

criteria for assessing the risk 

significance of the risk achievement 

worth values of common cause failures 

as part of the probabilistic risk 

assessment analysis of the risk 

importance of components. 

Date of issuance: August 12, 2009 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: Unit 1–191; Unit 

2–179. 

Facility Operating License Nos. NPF– 

76 and NPF–80: The amendments 

revised the Facility Operating Licenses, 

and Updated Final Safety Analysis 

Report. 

Date of initial notice in Federal 

Register: December 2, 2008 (73 FR 

73354). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 12, 

2009. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units 1 and 2, 

Louisa County, Virginia 

Date of application for amendments: 

February 6, 2009. 

Brief description of amendments: The 

proposed amendments deleted 

applicable portions of the Technical 

Specifications (TSs) superseded by Part 

26, Subpart I of Title 10 of the Code of 

Federal Regulations (10 CFR). This 

change is consistent with Nuclear 

Regulatory Commission (NRC)-approved 

Revision 0 to Technical Specification 

Task Force (TSTF) Improved Standard 

Technical Specification Change 

Traveler, TSTF–511, ‘‘Eliminate 

Working Hour Restrictions from TS 5.2– 

2 to Support Compliance with 10 CFR 

Part 26.’’ 

Date of issuance: July 29, 2009. 

Effective date: As of the date of 

issuance and shall be implemented by 

October 1, 2009. 

Amendment Nos.: 256 and 237. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

change the license and the technical 

specifications. 

Date of initial notice in Federal 

Register: March 24, 2009 (74 FR 

12396). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 29, 2009. No significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: August 

14, 2008. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3.3.2, ‘‘Engineered 

Safety Feature Actuation System 

(ESFAS) Instrumentation,’’ to extend the 

Surveillance Frequency on selected 

ESFAS slave relays from 92 days to 18 

months. 

Date of issuance: July 30, 2009. Effective date: As of its date of 

issuance and shall be implemented 

within 90 days of the date of issuance. Amendment No.: 183. 

Renewed Facility Operating License 

No. NPF–42. The amendment revised 

the Operating License and Technical 

Specifications. Date of initial notice in Federal 

Register: October 7, 2008 (73 FR 

58379). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 30, 2009. No significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: August 

14, 2008, as supplemented by letter 

dated April 10, 2009. Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3.3.2, ‘‘Engineered 

Safety Feature Actuation System 

(ESFAS) Instrumentation,’’ TS 3.7.2, 

‘‘Main Steam Isolation Valves (MSIVs),’’ 

and added new TS 3.7.19, ‘‘Secondary 

System Isolation Valves (SSIVs).’’ TS 

3.7.2 has been revised to add MSIV 

bypass valves to the scope of TS 3.7.2. 

TS Table 3.3.2–1 has been revised to 

reflect the addition of the MSIV bypass 

valves to TS 3.7.2 and the associated 

applicability to be consistent with 

Westinghouse Standard Technical 

Specifications (NUREG–1431, Revision 

3.0). TS 3.7.19 has been added to 

include a limiting condition for 

operation, conditions/required actions, 

and surveillance requirements for the 

steam generator blowdown isolation 

valves and steam generator blowdown 

sample isolation valves. 


pwalker on DSK8KYBLC1PROD with NOTICES
42940 Federal Register / Vol. 74, No. 163 / Tuesday, August 25, 2009 / Notices 

Date of issuance: July 31, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to startup from Refueling Outage 

17. 

Amendment No.: 184. 

Renewed Facility Operating License 

No.: NPF–42. The amendment revised 

the Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 7, 2008 (73 FR 

58679). The supplemental letter dated 

April 10, 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 July 31, 2009. 

No significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: March 6, 

2009, as supplemented by letter dated 

July 14, 2009. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 5.2.2, ‘‘Unit Staff,’’ to 

eliminate working hour restrictions (TS 

5.2.2.d) to support compliance with 

Title 10 of the Code of Federal 

Regulations (10 CFR) Part 26. In 

addition, paragraphs e and f of TS 5.2.2 

were renumbered to d and e to reflect 

the removal of paragraph d of TS 5.2.2, 

and a reference in 5.2.2b was updated 

to reflect the renumbering of 5.2.2f. to 

5.2.2e. The request is consistent with 

the guidance contained in U.S. Nuclear 

Regulatory Commission (NRC)-approved 

TS Task Force (TSTF) 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: August 7, 2009. 

Effective date: As of its date of 

issuance and shall be implemented by 

October 1, 2009. 

Amendment No.: 185. 

Renewed Facility Operating License 

No.: NPF–42. The amendment revised 

the Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 21, 2009 (74 FR 18258). 

The supplemental letter dated July 14, 

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 August 7, 2009. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 14th day 

of August 2009. 

For the Nuclear Regulatory Commission. 

Allen G. Howe, 

Acting Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–20403 Filed 8–24–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0371; Docket No. 030–14680] 

Notice of Availability of Environmental 

Assessment and Finding of No 

Significant Impact for License 

Amendment to Byproduct Materials 

License No. 29–00117–06, for 

Unrestricted Release of the Merck and 

Company’s Facility in Rahway, NJ 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Issuance of Environmental 

Assessment and Finding of No 

Significant Impact for License 

Amendment. 

FOR FURTHER INFORMATION CONTACT: 

Betsy Ullrich, Senior Health Physicist, 

Commercial & R&D Branch, Division of 

Nuclear Materials Safety, Region I, 475 

Allendale Road, King of Prussia, PA 

19406; telephone (610) 337–5040; fax 

number (610) 337–5269; or by e-mail: 

Elizabeth.ullrich@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The U.S. Nuclear Regulatory 

Commission (NRC) is considering the 

issuance of a license amendment to 

Byproduct Materials License No. 29– 

00117–06. This license is held by Merck 

and Company, Inc (the Licensee), for its 

Merck and Company, Merck Research 

Laboratories (the Facility), located at 

126 East Lincoln Avenue in Rahway, 

New Jersey. Issuance of the amendment 

would authorize release of the Facility’s 

Waste Incinerator for unrestricted use. 

The Licensee requested this action in a 

letter dated May 21, 2009. The NRC has 

prepared an Environmental Assessment 

(EA) in support of this proposed action 

in accordance with the requirements of 

Title 10, Code of Federal Regulations 

(CFR), part 51 (10 CFR part 51). Based 

on the EA, the NRC has concluded that 

a Finding of No Significant Impact 

(FONSI) is appropriate with respect to 

the proposed action. The amendment 

will be issued to the Licensee following 

the publication of this FONSI and EA in 

the Federal Register. 

II. Environmental Assessment 

Identification of Proposed Action 

The proposed action would approve 

the Licensee’s May 21, 2009 license 

amendment request, resulting in release 

of the Waste Incinerator for unrestricted 

use. License No. 29–00117–06 was 

issued on August 11, 1978, pursuant to 

10 CFR part 30, and has been amended 

periodically since that time. This 

license authorizes the Licensee to use 

unsealed byproduct material for 

purposes of conducting research and 

development activities on laboratory 

bench tops and in hoods, and 

incineration of radioactive waste. 

The Waste Incinerator is situated 

within Building 77 at 126 East Lincoln 

Avenue, and consists of the incinerator 

room and associated effluent component 

parts and mechanical component parts. 

The Waste Incinerator is located in an 

industrial area. Within the Waste 

Incinerator, use of licensed materials 

was confined to the Conveyor System 

Area, the Cold Room Area, the Burn 

Chamber and Kiln Area, the Loading 

Ram Area, the Loading Dock Area, the 

Fly Ash System and Bag House Area, 

the Restroom, the Mechanical Room, 

and the Control Room and its Stairwell. 

In 2009, the Licensee ceased using the 

Waste Incinerator for licensed waste 

disposal and initiated a survey and 

decontamination of the Waste 

Incinerator. Based on the Licensee’s 

historical knowledge of the site and the 

conditions of the Waste Incinerator, the 

Licensee determined that only routine 

decontamination activities, in 

accordance with their NRC-approved, 

operating radiation safety procedures, 

were required. The Licensee was not 

required to submit a decommissioning 

plan to the NRC because worker cleanup 

activities and procedures are consistent 

with those approved for routine 

operations. The Licensee conducted 

surveys of the Waste Incinerator and 

provided information to the NRC to 

demonstrate that it meets the criteria in 

subpart E of 10 CFR part 20 for 

unrestricted release. 

Need for the Proposed Action 

The Licensee has ceased using the 

Waste Incinerator for disposal of 

licensed materials at the Facility and 


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