Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 76407-76420 [E8-29450]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 242 (Tuesday, December 16, 2008)]
[Notices]
[Pages 76407-76420]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-29450]


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NUCLEAR REGULATORY COMMISSION


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 staff) 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

[[Page 76408]]

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 November 20, 2008 to December 3, 2008. The 
last biweekly notice was published on December 2, 2008 (73 FR 73351).

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 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, 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. The filing of requests for 
a hearing and petitions for leave to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. 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 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available documents 
related to these actions 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, 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 set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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.

[[Page 76409]]

    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated on August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve 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 a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@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 NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer \TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available 
at https://www.nrc.gov/site-help/e-submittals/install-viewer.html. 
Information about applying for a digital ID certificate is available on 
NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/
apply-certificates.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 https://www.nrc.gov/
site-help/e-submittals.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 may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, 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 petition 
and/or request 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). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://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. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, 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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
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 amendments request: October 1, 2008.
    Description of amendments request: The proposed amendment would 
insert a requirement into the operating licenses of the Calvert Cliffs 
Nuclear Power Plant, Unit Nos. 1 and 2, involving the reporting of 
specified reactor vessel (RV) inservice inspection (ISI) information 
and analyses as specified in Federal Register Notice (72 FR 56275), 
dated October 3, 2007, ``Alternative Fracture Toughness Requirements 
for Protection Against Pressurized Thermal Shock Events.'' This 
amendment is a required part of a code relief request, submitted by the 
licensee on October 1, 2008, to extend the RV ISI 10-year inspection 
interval for RV weld examinations.
    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?
    The proposed change, which adds a requirement within Calvert 
Cliffs licenses to provide required information and analyses as

[[Page 76410]]

a supporting condition for extending the allowed reactor vessel ISI 
interval, only involves the commitment to provide data obtained from 
the reactor vessel ISI. This proposed change involves only the 
submittal of generated data that will be used to verify the reactor 
vessel has more than sufficient margin to prevent any pressurized 
thermal shock event from occurring. This proposed change does not 
involve any change to the design basis of the plant or of any 
structure, system, or component. Therefore, the proposed change does 
not involve a significant increase in the probability or consequence 
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?
    The proposed change, which adds a requirement within Calvert 
Cliffs licenses to provide required information and analyses as a 
supporting condition for extending the reactor vessel ISI interval, 
only involves the commitment to provide data and analyses obtained 
from the reactor vessel ISI. As such this proposed change does not 
result in physical alteration to the plant configuration or make any 
change to plant operation. As a result no new accident scenarios, 
failure mechanisms, or single failures are introduced. 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 proposed change, which adds a requirement within Calvert 
Cliffs licenses, to provide required information and analyses as a 
supporting condition for extending the allowed reactor vessel ISI 
interval, only involves the commitment to provide data and analyses 
obtained from the reactor vessel ISI. The submitted data may be used 
to verify the condition of the reactor vessel meets all required 
standards to ensure sufficient safety margin is maintained against 
the occurrence of a pressurized thermal shock event during the 
expanded time interval between reactor vessel ISIs. The proposed 
change is administrative in nature and is not related to any margin 
[of] safety. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 18, 2008.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements for inoperable 
snubbers by relocating the current TS 3.7.8, ``Snubbers,'' to the 
Technical Requirements Manual (TRM) and adding Limiting Condition for 
Operation (LCO) 3.0.8. The proposed amendment would also make 
conforming changes to TS LCO 3.0.1. In conjunction with the proposed 
changes, the TS Bases for LCO 3.0.8 will be added, consistent with 
Bases Control Program, as described in Section 6.16 of the TS.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on November 24, 2004 (69 FR 68412), on possible 
license amendments adopting TSTF-372 using the NRC's CLIIP for amending 
licensee's TSs, which included a model safety evaluation (SE) and model 
no significant hazards consideration (NSHC) determination.
    The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on May 4, 2005. (70 FR 23252), which included the resolution 
of public comments on the model SE. The May 4, 2005, notice of 
availability referenced the November 4, 2004, notice. The licensee has 
affirmed the applicability of the following NSHC determination in its 
application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change[s] [Do] Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change[s] [allow] a delay time for entering a 
supported system TS when the inoperability is due solely to an 
inoperable snubber if risk is assessed and managed. The postulated 
seismic event requiring snubbers is a low-probability occurrence and 
the overall TS system safety function would still be available for the 
vast majority of anticipated challenges. Therefore, the probability of 
an accident previously evaluated is not significantly increased, if at 
all. The consequences of an accident while relying on allowance 
provided by proposed LCO 3.0.8 are no different than the consequences 
of an accident while relying on the TS required actions in effect 
without the allowance provided by proposed LCO 3.0.8. Therefore, the 
consequences of an accident previously evaluated are not significantly 
affected by [these] change[s]. The addition of a requirement to assess 
and manage the risk introduced by [these] change[s] will further 
minimize possible concerns. Therefore, [these] change[s] [do] not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
Criterion 2--The Proposed Change[s] [Do] Not Create the Possibility of 
a New or Different Kind of Accident From Any Previously Evaluated
    The proposed change[s] [do] not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is assessed 
and managed, will not introduce new failure modes or effects and will 
not, in the absence of other unrelated failures, lead to an accident 
whose consequences exceed the consequences of accidents previously 
evaluated. The addition of a requirement to assess and manage the risk 
introduced by [these] change[s] will further minimize possible 
concerns. Thus, [these] change[s] [do] not create the possibility of a 
new or different kind of accident from an accident previously 
evaluated.
Criterion 3--The Proposed Change[s] [Do] Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change[s] [allow] a delay time for entering a 
supported system TS when the inoperability is due solely to an 
inoperable snubber, if risk is assessed and managed. The postulated 
seismic event requiring snubbers is a low-probability occurrence and 
the overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three tiered approach 
recommended in NRC Regulatory Guide 1.177. A bounding risk assessment 
was performed to justify the proposed TS changes. This application of 
LCO 3.0.8 is predicated upon the licensee's performance of a risk 
assessment and the management of plant risk. The net change to the 
margin of safety is insignificant. Therefore, [these] change[s] [do] 
not involve a significant reduction in a margin of safety.

[[Page 76411]]

    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.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County, 
Pennsylvania

    Date of amendment request: November 7, 2008.
    Description of amendment request: The proposed amendment would 
modify the method used to calculate the available net positive suction 
head (NPSH) for the BVPS-2 recirculation spray (RS) pumps as described 
in the BVPS-2 Updated Final Safety Analysis Report (UFSAR). BVPS-2 
UFSAR would take credit for containment overpressure by allowing for 
the difference between containment total pressure and the vapor 
pressure of the water in the containment sump in the available NPSH 
calculation.
    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 change to the method used to calculate available NPSH for 
the RS pumps will not affect the probability of an accident because 
the RS pumps are not used during normal plant operations and cannot 
initiate an accident.
    Successful operation of at least one train of RS pumps is 
required in order to demonstrate that containment and fuel cladding 
design basis limits are not exceeded. The design basis accident 
currently assumes a breach of the reactor coolant pressure boundary. 
There is no impact to the fuel cladding since the proposed change 
does not affect performance of the emergency core cooling systems. 
Successful operation of the RS pumps depends on adequate NPSH being 
available to support RS pump performance. The change in the 
methodology will result in an increase of the NPSH available to the 
RS pumps as calculated in the safety analysis. This will increase 
the calculated NPSH margin because the required NPSH to the RS pumps 
will not change due to the methodology change. Because the available 
NPSH remains adequate, with margin to NPSH requirements, acceptable 
RS pump performance will be assured and the design basis limits for 
containment pressure and fuel cladding will not be exceeded and the 
consequences of an accident will not be increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The change to the method used to calculate available NPSH for 
the RS pumps will not create the possibility of a new accident 
because the operation of the plant or the RS pumps is not changed. 
The RS pumps are not used during normal plant operations and cannot 
initiate an accident. A different kind of accident will not be 
created because the proposed calculation method will produce an NPSH 
value that will ensure proper operation of the pumps and will not 
result in any new failure modes of the RS pumps.
    Therefore, the proposed amendment will 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 change to the method used to calculate available NPSH for 
the RS pumps will not involve a significant reduction in a margin of 
safety because the change does not reduce the NPSH margin to the RS 
pump required NPSH. The only controlling numerical value pertaining 
to available NPSH of the RS pumps that is established in the UFSAR 
is a lower limit specified in the UFSAR, referred to as the required 
NPSH for the RS pumps. The required NPSH limit will not be altered 
as a result of the proposed calculation method, and the required 
NPSH will continue to be maintained under the applicable accident 
scenario.
    Therefore, the proposed amendment will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Nuclear Operating Company, FirstEnergy Corporation, 76 South Main 
Street, Akron, OH 44308.
    NRC Branch Chief: Mark G. Kowal.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: September 25, 2008.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications, Figures 4.3-1 and 4.3-2, which show 
allowable locations for nuclear fuel in the spent fuel pool storage 
racks. The figures currently show two different allowable storage 
patterns for four of the storage rack modules. I&M proposes to modify 
these two figures such that fuel may be located in any of these four 
individual modules in accordance with either figure to allow continued 
placement of new and intermediate burn-up fuel in the spent fuel pool 
as the storage racks approach capacity.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has performed its own analysis, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The accidents and events of concern involving fuel located in 
the spent fuel pool storage racks are a criticality accident, a fuel 
handling accident, and inadequate decay heat removal. The proposed 
change will not increase the probability of a criticality accident 
because analyses demonstrate that sub-criticality will be maintained 
for the fuel storage considerations allowed by the change. The 
proposed change will not increase the probability of a fuel handling 
accident because it does not affect the manner in which fuel is 
moved or handled. The proposed change will decrease the number of 
fuel moves needed for upcoming refueling outages. The proposed 
change will not increase the probability of inadequate decay heat 
removal because thermal-hydraulic analyses demonstrate adequate heat 
removal will remain valid for the storage configurations allowed by 
the change. Therefore, the probability of occurrence of a previously 
evaluated accident will not be significantly increased.
    The proposed change does not adversely affect the ability to 
perform the intended safety functions of any structure, system, or 
component (SSC) credited for mitigating a criticality accident, a 
fuel handling accident, or inadequate decay heat removal. Therefore, 
the consequences of a previously evaluated accident will not be 
significantly increased.
    Therefore, the proposed amendment 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 does not alter the design function or 
operation of any SSC. The proposed change does not affect the 
capability of the SSCs involved with the storage of fuel in the 
spent fuel pool to

[[Page 76412]]

perform their function. As a result, no new failure mechanisms, 
malfunctions, or accident initiators are created. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margins of safety involved with the storage of fuel in the 
spent fuel pool are the margins associated with criticality, 
mitigation of a fuel handling accident, and assurance of adequate 
decay heat removal. The proposed amendment involves no change in the 
capability of any SSC that maintains these margins. Therefore, there 
is no significant reduction in a margin of safety as a result of the 
proposed amendment.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on its own analysis, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Lois M. James.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: October 21, 2008.
    Description of amendment request: The proposed amendment would 
modify Technical Specification 5.6.3, ``Radioactive Effluent Release 
Report,'' by changing the required annual submittal date for the report 
from ``within 90 days of January 1'' (i.e., prior to April 1), to prior 
to May 1. The change is consistent with the requirements for the 
Radioactive Effluent Release Report submittal date identified in 
Technical Specification Task Force Traveler Number 152 (TSTF-152), 
``Revise Reporting Requirements to be Consistent with 10 CFR 20,'' 
approved by the U.S. Nuclear Regulatory Commission (NRC) in March 1997.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has performed its own analysis, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change is administrative in nature. The date of the 
submittal of the Radioactive Effluent Release Report is not an 
initiator of any analyzed event. Similarly, the date of submission 
does not affect the consequences of any accident previously 
evaluated. The proposed change does not physically alter the plant 
or affect plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in nature. It revises the 
date by which the Radioactive Effluent Release Report is required to 
be submitted to the NRDC. Revision of the submittal date of the 
report does not affect any accident initiator or cause any new 
accident precursors to be created. The proposed change does not 
affect the types or amounts of radioactive effluents released or 
cumulative occupational radiological exposures.
    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 is administrative in nature and does not 
involve a significant reduction in a margin of safety. There are no 
margins of safety associated with the submittal date for the 
Radioactive Effluent Release Report.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on its own analysis, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Lois M. James.

Indiana Michigan Power Company (I&M), Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of amendment request: October 9, 2008.
    Description of amendment request: The proposed amendment would 
support a proposed change to the inservice inspection program that is 
based on topical report WCAP-16168-NP-A, Revision 2, ``Risk-Informed 
Extension of the Reactor Vessel Inservice Inspection Interval.'' The 
U.S. Nuclear Regulatory Commission (NRC) safety evaluation approving 
the topical report requires licensees to amend their licenses to 
require that the information and analyses requested in Section (e) of 
the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 
56275 prior to issuance of the final 10 CFR 50.61a) be submitted for 
NRC staff review and approval within 1 year of completing the required 
reactor vessel weld inspection. I&M proposes to add a new license 
condition to provide this information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise the license to require the 
submission of information and analyses to the Nuclear Regulatory 
Commission (NRC) following completion of each American Society of 
Mechanical Engineers (ASME) Code, Section XI, Category B-A and B-D 
Reactor Vessel weld inspection. Submittal of the information and 
analyses can have no effect on the consequences of an accident or 
the probability of an accident because the submission of information 
is not related to the operation of the plant or any equipment, the 
programs and procedures used to operate the plant, or the evaluation 
of accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will only affect the requirement to submit 
information and analyses when specified inspections are performed. 
There are no changes to plant equipment, operating characteristics 
or conditions, programs or failures. There are no new accident 
initiators or precursors.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will revise the license to require the 
submission of information and analyses to the NRC following 
completion of each ASME Code, Section XI, Category B-A and B-D 
Reactor Vessel weld inspection which does not affect any Limiting 
Conditions for Operation used to establish the margin of safety. The 
requirement to submit information and analyses is an administrative 
tool to assure the NRC has the ability to independently review 
information developed by the licensee.

[[Page 76413]]

    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: James M. Petro, Jr., Senior Nuclear Counsel, 
Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Lois M. James.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: October 7, 2008.
    Description of amendment request: The proposed amendment would 
insert a requirement into the operating license of the Ginna Nuclear 
Power Plant involving the reporting of specified reactor vessel (RV) 
inservice inspection (ISI) information and analyses as specified in 
Federal Register Notice (72 FR 56275), dated October 3, 2007, 
``Alternative Fracture Toughness Requirements for Protection Against 
Pressurized Thermal Shock Events.'' This amendment is a required part 
of a code relief request, submitted by the licensee on October 3, 2008, 
to extend the RV ISI 10-year inspection interval.
    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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change, which adds a requirement within the Ginna 
license, to provide required information and analyses as a 
supporting condition for extending the allowed reactor vessel ISI 
interval, only involves the commitment to provide data obtained from 
the reactor vessel ISI. This proposed change involves only the 
submittal of generated data that will be used to verify the reactor 
vessel has more than sufficient margin to prevent any pressurized 
thermal shock event from occurring. This proposed change does not 
involve any change to the design basis of the plant or of any 
structure, system, or component. Therefore, the proposed change does 
not involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change, which adds a requirement within the Ginna 
license to provide required information and analyses as a supporting 
condition for extending the reactor vessel ISI interval, only 
involves the commitment to provide data and analyses obtained from 
the reactor vessel ISI. As such this proposed change does not result 
in physical alteration to the plant configuration or make any change 
to plant operation. As a result no new accident scenarios, failure 
mechanisms, or single-failures are introduced. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed change, which adds a requirement within the Ginna 
license, to provide required information and analyses as a 
supporting condition for extending the allowed reactor vessel ISI 
interval, only involves the commitment to provide data and analyses 
obtained from the reactor vessel ISI. The submitted data will be 
used to verify the condition of the reactor vessel meets all 
required standards to ensure a sufficient safety margin is 
maintained against the occurrence of a pressurized thermal shock 
event during the expanded time interval between reactor vessel ISIs. 
The proposed change is administrative in nature and is not related 
to any margin to safety. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, 
Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: October 8, 2008.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) by the adoption of Technical 
Specification Task Force (TSTF) Standard TS Change Traveler TSTF-374, 
Revision 0, to modify TS by relocating references to specific American 
Society for Testing and Materials (ASTM) standards for fuel oil testing 
to licensee-controlled documents and adding alternate criteria to the 
``clear and bright'' acceptance test for new fuel oil. The proposed 
change was described in the Notice of Availability published in the 
Federal Register on April 21, 2006 (71 FR 20735).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (NSHC) by incorporating by reference the proposed NSHC 
determination (NSHCD) presented in the Federal Register notice on 
February 22, 2006 (71 FR 9179), which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Requirements to perform testing in 
accordance with applicable ASTM standards are retained in the TS as 
are requirements to perform surveillances of both new and stored 
diesel fuel oil. Future changes to the licensee-controlled document 
will be evaluated pursuant to the requirements of 10 CFR 50.59, 
``Changes, tests and experiments,'' to ensure that such changes do 
not result in more than a minimal increase in the probability or 
consequences of an accident previously evaluated. In addition, the 
``clear and bright'' test used to establish the acceptability of new 
fuel oil for use prior to addition to storage tanks has been 
expanded to recognize more rigorous testing of water and sediment 
content. Relocating the specific ASTM standard references from the 
TS to a licensee-controlled document and allowing a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil will not affect nor degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.

[[Page 76414]]

    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any 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 changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. The requirements retained in the TS continue to require 
testing of the diesel fuel oil to ensure the proper functioning of 
the DGs.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of applicable ASTM standards to evaluate 
the quality of both new and stored fuel oil designated for use in 
the emergency DGs. Changes to the licensee-controlled document are 
performed in accordance with the provisions of 10 CFR 50.59. This 
approach provides an effective level of regulatory control and 
ensures that diesel fuel oil testing is conducted such that there is 
no significant reduction in a margin of safety.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality for emergency DG use. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
preserve the current margins 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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Branch Chief: Melanie Wong.

Virginia Electric and Power Company, Docket No. 50-280, Surry Power 
Station, Unit No. 1, Surry County, Virginia

    Date of amendment request: October 14, 2008.
    Description of amendment request: The proposed change includes a 
one-cycle revision to the Surry Power Station, Unit No. 1 (Surry 1) 
technical specifications (TSs). Specifically, TS 6.4.Q, ``Steam 
Generator (SG) Program,'' and TS 6.6.A.3, ``Steam Generator Tube 
Inspection Report,'' will be revised to incorporate an interim 
alternate repair criterion into the provisions for SG tube repair for 
use during the Surry 1 2009 spring refueling outage and the subsequent 
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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Of the various accidents previously evaluated, the proposed 
changes only affect the steam generator tube rupture (SGTR) event 
evaluation and the postulated steam line break (SLB), and locked 
rotor evaluations. Loss-of-coolant accident (LOCA) conditions cause 
a compressive axial load to act on the tube. Therefore, since the 
LOCA tends to force the tube into the tubesheet rather than pull it 
out, it is not a factor in this amendment request.
    Another faulted load consideration is a safe shutdown earthquake 
(SSE); however, the seismic analysis of Model F steam generators has 
shown that axial loading of the tubes is negligible during an SSE. 
At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) below 17 inches from the TTS [top of the 
tubesheet] is limited by both the tube-to-tubesheet crevice and the 
limited crack opening, permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region.
    For the SGTR event, the required structural margins of the steam 
generator tubes is maintained by limiting the allowable ligament 
size for a circumferential crack to remain in service to 203 degrees 
below 17 inches from the TTS for the subsequent operating cycle. 
Tube rupture is precluded for cracks in the hydraulic expansion 
region due to the constraint provided by the tubesheet. The 
potential for tube pullout is mitigated by limiting the allowable 
crack size to 203 degrees for the subsequent operating cycle. These 
allowable crack sizes take into account eddy current uncertainty and 
crack growth rate. It has been shown that a circumferential crack 
with an azimuthal extent of 203 degrees for the 18 month SG tubing 
eddy current inspection interval meet the performance criteria of 
NEI 97-06, Rev. 2, ``Steam Generator Program Guidelines'' and 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes.'' Therefore, the margin against tube burst/pullout 
is maintained during normal and postulated accident conditions and 
the proposed change does not result in a significant increase in the 
probability or consequence of a SGTR.
    The probability of a SLB is unaffected by the potential failure 
of a SG tube as the failure of a tube is not an initiator for a SLB 
event. SLB leakage is limited by leakage flow restrictions resulting 
from the leakage path above potential cracks through the tube-to-
tubesheet crevice. The leak rate during postulated accident 
conditions (including locked rotor) has been shown to remain within 
the accident analysis assumptions for all axial or circumferentially 
oriented cracks occurring 17 inches below the top of the tubesheet. 
Since normal operating leakage is limited to 150 gpd [gallons per 
day], the attendant accident condition leak rate, assuming all 
leakage to be from indications below 17 inches from the top of the 
tubesheet, would be bounded by 470 gpd. This value is within the 
accident analysis assumptions for the limiting design basis accident 
for Surry, which is the postulated SLB event.
    Based on the above, the performance criteria of NEI-97-06, Rev. 
2 and Regulatory Guide (RG) 1.121 continue to be met and 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 introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon implementation of the interim alternate repair 
criteria. The proposed change does not introduce any new equipment 
or any change to existing equipment. No new effects on existing 
equipment are created nor are any new malfunctions introduced.
    Therefore, based on the above evaluation, 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change maintains the required structural margins of 
the steam generator tubes for both normal and accident conditions. 
NEI 97-06, Rev. 2 and RG 1.121 are used as the basis in the 
development of the limited tubesheet inspection depth methodology 
for determining that steam generator tube integrity considerations 
are maintained within acceptable limits. RG

[[Page 76415]]

1.121 describes a method acceptable to the NRC staff for meeting GDC 
14, 15, 31, and 32 by reducing the probability and consequences of 
an SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions of tube wall degradation beyond which tubes with 
unacceptable cracking, as established by inservice inspection, 
should be removed from service or repaired, the probability and 
consequences of a SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the ASME Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking in a tube or the tube-to-
tubesheet weld, References 2 and 4 [of the application] define a 
length of remaining tube ligament that provides the necessary 
resistance to tube pullout due to the pressure induced forces (with 
applicable safety factors applied). Additionally, it is shown that 
application of the limited tubesheet inspection depth criteria will 
not result in unacceptable primary-to-secondary leakage during all 
plant conditions.
    Based on the above, it is concluded that the proposed changes do 
not result in any reduction of margin with respect to plant safety 
as defined in the Updated Final Safety Analysis Report or bases of 
the plant Technical Specifications.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, 
VA 23219.
    NRC Branch Chief: Melanie C. Wong.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: October 9, 2008.
    Description of amendment request: The proposed change revises the 
technical specifications (TSs) for consistency with the assumptions of 
the current Alternate Source Term dose analysis of record, performed in 
accordance with Title 10 of the Code of Federal Regulations (10 CFR), 
Section 50.67, and the results of non-pressurized main control room/
emergency switchgear room (MCR/ESGR) envelope boundary tracer gas 
testing. The proposed change removes the MCR Bottled Air System 
requirements from the 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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The MCR Bottled Air System is not an initiator or precursor 
to any accident previously evaluated, and is not credited as a 
success path for dose mitigation in the event of a DBA [design-basis 
accident]. MCR/ESGR envelope isolation and emergency ventilation 
continue to be available consistent with accident analyses 
assumptions. Therefore, the proposed TS 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?
    Response: No.
    The proposed change does not alter the requirements for MCR/ESGR 
envelope isolation or the MCR/ESGR Emergency Ventilation System 
during accident conditions. No physical modifications to the plant 
are being made (i.e., no new or different type of equipment will be 
installed), and no significant changes in the methods governing 
normal plant operation are being implemented. Also, the proposed 
change does not alter assumptions made in the safety analysis and is 
consistent with those assumptions. Therefore, the proposed TS change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined, and the dose analysis acceptance criteria 
are not affected. The proposed change does not result in plant 
operation in a configuration outside the analyses or design basis 
and does not adversely affect systems that respond to safely shut 
down the plant and to maintain the plant in a safe shutdown 
condition. Therefore, the proposed TS change does not involve a 
significant reduction in a 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, 
VA 23219.
    NRC Branch Chief: Melanie C. Wong.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 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, 
                    

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