Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 11383-11403 [E7-4251]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 48 (Tuesday, March 13, 2007)]
[Notices]
[Pages 11383-11403]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-4251]


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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 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 February 15, 2007 through March 1, 2007. 
The last biweekly notice was published on February 27, 2007 (72 FR 
8800).

[[Page 11384]]

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. Within 60 days after the date of publication of this 
notice, the licensee 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 for a hearing and a petition for leave to intervene.
    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. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. 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, the 
licensee 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 
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 persons 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 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, 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

[[Page 11385]]

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.
    A request for a hearing or a petition for leave to intervene must 
be filed 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; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this 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 amendment request: February 1, 2007.
    Description of amendment request: The proposed license amendment 
would revise Surveillance Requirement (SR) 3.5.2.8 in Technical 
Specification 3.5.2, ``ECCS [Emergency Core Cooling System]--
Operating,'' to reflect the replacement of the containment 
recirculation sump suction inlet trash racks and screens with 
strainers, in response to Nuclear Regulatory Commission (NRC) Generic 
Letter 2004-02, ``Potential Impact of Debris Blockage on Emergency 
Recirculation during Design Basis Accidents at Pressurized-Water 
Reactors.'' The proposed license amendment would replace ``trash racks 
and screens'' with ``strainers'' in SR 3.5.2.8.
    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?
    The consequences of accidents evaluated in the Updated Final 
Safety Analysis Report [UFSAR] that could be affected by the 
proposed change are those involving the pressurization of 
Containment and associated flooding of the Containment and 
recirculation of this fluid within the Emergency Core Cooling System 
(ECCS) or the Containment Spray System (CSS) (e.g., loss-of-coolant 
accidents [LOCAs]). The proposed change does not impact the 
initiation or probability of occurrence of any accident. Although 
the configurations of the existing containment recirculation sump 
trash racks and screen and the replacement sump strainer cassettes 
are different, they serve the same fundamental purpose of passively 
removing debris from the sump's suction supply of the supported 
system pumps. Removal of trash racks does not impact the adequacy of 
the pump net positive suction head assumed in the safety analysis. 
Likewise, the change does not reduce the reliability of any 
supported systems or introduce any new system interactions. The 
greatly increased surface area of the new strainer is designed to 
reduce head loss and reduce the approach velocity at the strainer 
face significantly, decreasing the risk of impact from large debris 
entrained in the sump flow stream.
    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?
    The containment recirculation sump strainers are a passive 
system used for accident mitigation. As such, they cannot be 
accident initiators. Therefore, there is no possibility that this 
change could create any new or different kind of accident. No new 
accident scenarios, transient precursors, or limiting single 
failures are introduced as a result of the proposed change. There 
will be no adverse effect or challenges imposed on any safety-
related system as a result of the change. Therefore, the possibility 
of a new or different [kind] of accident is not created.
    There are no changes which would cause the malfunction of 
safety-related equipment, assumed to be OPERABLE in the accident 
analyses, as a result of the proposed Technical Specification 
change. No new equipment performance burdens are imposed. The 
possibility of a malfunction of safety-related equipment with a 
different result is not created.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind of] accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any safety analysis 
limit. There will be no effect on the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. The proposed change does not adversely affect the fuel, 
fuel cladding, Reactor Coolant System, or containment integrity. The 
radiological dose consequence acceptance criteria listed in the 
Updated Final Safety Analysis Report will continue to be met.
    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, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt 
Street, 17th floor, Baltimore, MD 21202.
    NRC Acting Branch Chief: John P. Boska.

Carolina Power & Light Company, Docket Nos. 50-325 and 5-324 Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

    Date of amendments request: December 21, 2006.
    Description of amendment request: The proposed amendment would

[[Page 11386]]

modify technical specification (TS) requirements of TS 3.4.1, 
``Recirculation Loops Operating,'' to require the recirculation loops 
be operated with matched flows versus recirculation pump speeds as 
currently required. This change affects the Limiting Condition for 
Operation (LCO) requirements and Surveillance Requirements (SRs) of TS 
3.4.1.
    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 amendment implements more conservative requirements 
associated with recirculation loop operation. Specifically, the LCO 
requirements of TS 3.4.1 and SR 3.4.1.1 are being revised to 
directly monitor recirculation loop jet pump flows versus 
recirculation pump speed, eliminating potential non-conservatism 
associated with relating recirculation loop jet pump flow to 
recirculation pump speed. These requirements assure that the 
mismatch between recirculation loop jet pump flows are bounded by 
the existing design bases analyses. As a result, the proposed change 
ensures that the consequences of a design bases LOCA [loss-of-
coolant accident] remain within the existing evaluation.
    The proposed change does not involve a physical change to the 
Reactor Recirculation system, nor does it alter the assumptions of 
the accident analyses. Therefore the probability of an accident 
previously evaluated is not affected.
    Based on the above, 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 does not involve a physical change to the 
Reactor Recirculation system, nor does it alter the assumptions of 
the accident analyses.
    The implementation of more conservative requirements associated 
with recirculation loop operation does not introduce any new failure 
modes. 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 amendment implements more conservative requirements 
associated with recirculation loop operation. These requirements 
ensure that the Reactor Recirculation system is operated consistent 
with the initial conditions of the existing design bases analyses. 
Since the design bases analyses assumptions are unchanged, the 
proposed change does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: L. Raghavan.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: December 15, 2006.
    Description of amendment request: The amendment would incorporate 
changes to the Technical Specifications (TS) associated with previously 
approved industry initiatives. The first change would relocate the 
Safety Limit Violation specifications from the administrative controls 
TS section to the safety limit TS sections as approved by TSTF-05-A, 
``Deletion of Safety Limit Violation Requirements.'' The second change 
would incorporate generic position titles, as approved by TSTF-65-A, 
``Use of Generic Titles for Utility Positions,'' and incorporates 
changes approved by NRC Administrative Letter (AL) 95-06, ``Relocation 
of Technical Specification Administrative Controls Related to Quality 
Assurance.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment consists of changes to and relocation 
of administrative TS requirements that were previously generically 
approved by the NRC. The proposed amendment would not change any of 
the previously evaluated accidents in the updated safety analysis 
report (USAR). The administrative controls that are affected by the 
proposed amendment do not have any function related to preventing or 
mitigating any of these previously evaluated accidents. The proposed 
amendment does not affect any systems, structures, or components 
(SSCs) that have the function of preventing or mitigating any of 
these previously evaluated accidents. The proposed amendment does 
not increase the likelihood of the malfunction of an SSC, thus the 
potential impact on analyzed accidents need not be considered.
    Because the proposed amendment is a relocation of administrative 
requirements that are not associated with preventing or mitigating 
the consequences of any previously evaluated accidents, there is no 
affect on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed amendment consists of changes to and relocation 
of administrative TS requirements previously generically approved by 
the NRC. This amendment will not change the design function of any 
SSC or the manner that any SSC is operated. Because this amendment 
does not change the design function or operation of any SSC, the 
amendment would not create the possibility of a new or different 
kind of accident due to credible new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment consists of changes to and relocation 
of administrative TS requirements previously generically approved by 
the NRC. The amendment does not alter any design basis safety limit 
and no safety margins are affected.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Branch Chief: P. Milano.

Duke Power Company LLC, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1 (Catawba), York County, South Carolina

    Date of amendment request: November 22, 2006.
    Description of amendment request: The amendment would revise the 
Catawba Unit 1 Facility Operating License (FOL) to provide for an 
extension of the time limit to complete the required modification to 
the Emergency Core Cooling System (ECCS) sump.

[[Page 11387]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed license amendment delineates a new Unit 1 FOL 
condition to implement a completion date associated with the ECCS 
sump strainer modification. The proposed license amendment is 
administrative in nature and is being submitted to fulfill a 
commitment made in previous Duke licensing correspondence. 
Therefore, the proposed license amendment has no effect upon either 
the probability or consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    As stated above, the proposed license amendment is 
administrative in nature and does not change the manner in which 
Unit 1 is designed or operated. Therefore, the proposed license 
amendment cannot create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their intended functions. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment. The performance of these barriers will 
not be affected by the addition of the proposed FOL condition. Being 
administrative in nature, the proposed license amendment therefore 
does not involve a significant reduction in any safety margin.

    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: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: April 11, 2006.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) related to the organizational 
description in TS 5.2.1
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided it's 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 NRC staff's review 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 proposed change revises an organizational description in TS 
5.2.1 to reflect the change of the title of the Vice President Nuclear 
Generation. The change is solely administrative in nature and has no 
impact on any accident probabilities or consequences. The change does 
not affect structures or components in the plant. The change has no 
affect on any accident previously evaluated. Therefore the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Accident Previously 
Evaluated
    There are no new accident causal mechanisms created as a result of 
this proposed change. No changes are being made to the plant that will 
introduce any new accident causal mechanisms. The change is solely 
administrative in nature and does not impact any plant systems that are 
accident initiators. Therefore, no new accidents or a different 
accident than previously evaluated is being created.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their design functions during and 
following an accident situation. The proposed change is solely 
administrative in nature and does not affect the performance of the 
barriers. Consequently, no safety margins will be impacted. 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: Ms. Lisa F. Vaughn, Duke Power Company LLC, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN 
50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 
2 and 3, Grundy County, Illinois.
    Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois.
    Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, 
Units 2 and 3, York and Lancaster Counties, Pennsylvania.
    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois.
    Date of amendment request: December 15, 2006.
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TSs) by replacing the term 
``plant-specific titles'' with ``generic titles'' in TS Section 
5.2.1.a, ensuring the TS description is consistent with the EGC Quality 
Assurance Topical Report (QATR). The proposed amendment will also 
revise the Peach Bottom TS Section 5.2.1.a, to replace the reference to 
the Updated Final Safety Analysis Report with reference to the EGC 
QATR. This will align the Peach Bottom TS wording with the rest of the 
EGC fleet.
    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 is a word replacement in TS 5.2.1, ``Onsite 
and Offsite Organizations.'' The proposed change involves no changes 
to plant systems or accident analyses. The proposed change is 
administrative in nature and, as such, does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients.
    Therefore, the proposed change does not involve any increase in 
the probability or

[[Page 11388]]

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.
    Creation of the possibility of a new or different kind of 
accident would require creating one or more new accident precursors. 
New accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed change does not involve a physical alteration of the 
plant, add any new equipment, or allow any existing equipment to be 
operated in a manner different from the present method of 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative in nature and has no 
impact on equipment design or method of operation. There are no 
changes being made to safety limits or safety system allowable 
values that would adversely affect plant safety as a result of the 
proposed change.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Michael L. Marshall, Jr.

Exelon Generation Company, LLC (EGC) Docket Nos. 50-254 and 50-265, 
Quad Cities Nuclear Power Station, Unit 1, Rock Island County, Illinois

    Date of amendment request: January 16, 2007.
    Description of amendment request: The proposed amendment revises 
the values of the safety limit minimum critical power ratio (SLMCPR) in 
the Quad Cities Nuclear Power Station (QCNPS), Unit 1, Technical 
Specification (TS) Section 2.1.1, ``Reactor Core SLs [Safety Limits].'' 
Specifically, the proposed change would require that for QCNPS, Unit 1, 
the minimum critical power ratio shall be greater than 1.11 for two 
recirculation loop operation, or greater than 1.13 for single 
recirculation loop operation. This change is needed to support the next 
cycle of operation for QCNPS, Unit 1.
    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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the SLMCPR for QCNPS, Unit 1, 
Cycle 20 such that the fuel is protected during normal operation and 
during plant transients or anticipated operational occurrences 
(AOOs).
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during plant transients or AOOs. 
Operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs) is met. Since the proposed change does not affect 
operability of plant systems designed to mitigate any consequences 
of accidents, the consequences of an accident previously evaluated 
are not expected to increase.
    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.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed change does not involve any plant configuration 
modifications or changes to allowable modes of operation. The 
proposed change to the SLMCPR assures that safety criteria are 
maintained for QCNPS, Unit 1, Cycle 20.
    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 SLMCPR provides a margin of safety by ensuring that at least 
99.9% of the fuel rods do not experience transition boiling during 
normal operation and AOOs if the MCPR limit is not violated. The 
proposed change will ensure the current level of fuel protection is 
maintained by continuing to ensure that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs if the MCPR limit is not violated. The proposed SLMCPR 
values were developed using NRC-approved methods. Additionally, 
operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that no more than 0.1% of the 
rods are expected to be in boiling transition if the MCPR limit is 
not violated) is met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based upon the above, EGC concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of no 
significant hazards consideration is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Michael L. Marshall, Jr.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: December 29, 2006.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.1.8, ``Scram Discharge Volume (SDV) Vent 
and Drain Valves,'' to allow a vent or drain line with one inoperable 
valve to be isolated instead of requiring the valve to be restored to 
operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments concerning the consolidated line item implement process 
(CLIIP), including a model safety evaluation and a model no significant 
hazards consideration

[[Page 11389]]

(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 April 15, 2003 (68 FR 18294), 
as part of the CLIIP. In its application dated December 29, 2006, the 
licensee affirmed the applicability of the following determination.
    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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead or requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDVs is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.
    2. The proposed change does not 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. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Michael L. Marshall, Jr.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Lake County, Ohio, and Docket Nos. 50-334 
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver 
County, Pennsylvania

    Date of amendment request: January 11, 2007.
    Description of amendment request: The proposed license amendments 
would modify technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The 
proposed license amendments also modify LCO 3.0.1 to incorporate the 
addition of LCO 3.0.8. This change is based on the TS Task Force (TSTF) 
Traveler, TSTF-372, Revision 4. A notice of availability for this TS 
improvement using the consolidated line item improvement process was 
published in the Federal Register on May 4, 2005.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing license amendment applications in the 
Federal Register on November 24, 2004 (69 FR 68412), and May 4, 2005 
(70 FR 23252). The licensee affirmed the applicability of the model 
NSHC determination in its application dated January 11, 2007.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed change allows a delay time for entering a supported 
system technical specification (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 this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated.
    The proposed change does 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 this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety.
    The proposed change allows 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 RG 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, this 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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Michael L. Marshall, Jr.

[[Page 11390]]

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: December 14, 2006.
    Description of amendment request: The proposed license amendment 
would revise the accident source term used in the NMP1 design basis 
radiological consequence analyses in accordance with 10 CFR 50.67. The 
revised accident source term replaces the current methodology that is 
based on TID-14844, ``Calculation of Distance Factors for Power and 
Test Reactor Sites,'' with the alternative source term (AST) 
methodology described in Regulatory Guide (RG) 1.183, ``Alternative 
Source Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors.'' The amendment request is for full implementation of the AST 
as described in RG 1.183, with the exception that TID-14844 will 
continue to be used as the radiation dose basis for equipment 
qualification and vital area access.
    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.
    Adoption of the AST and those plant systems affected by 
implementing AST do not initiate DBAs [design-basis accidents]. The 
AST does not affect the design or manner in which the facility is 
operated; rather, for postulated accidents, the AST is an input to 
calculations that evaluate the radiological consequences. The AST 
does not by itself affect the post-accident plant response or the 
actual pathway of the radiation released from the fuel. It does, 
however, better represent the physical characteristics of the 
release, so that appropriate mitigation techniques may be applied. 
Implementation of the AST has been incorporated in the analyses for 
the limiting DBAs at NMP1.
    The structures, systems and components affected by the proposed 
change mitigate the consequences of accidents after the accident has 
been initiated. Application of the AST does result in changes to 
NMP1 Updated Final Safety Analysis Report (UFSAR) functions (e.g., 
Liquid Poison system). As a condition of the application of AST, 
NMPNS is proposing to use the Liquid Poison system to control the 
suppression pool pH following a LOCA [loss-of-coolant accident]. The 
proposed changes also revise operability requirements for the 
secondary containment and certain post-accident filtration systems 
while handling irradiated fuel that has decayed for greater than 24 
hours and during core alterations. These changes have been included 
within the AST evaluations. These changes do not require any 
physical changes to the plant. As a result, the proposed changes do 
not involve a revision to the parameters or conditions that could 
contribute to the initiation of a DBA discussed in Chapter XV of the 
NMP1 UFSAR. Since design basis accident initiators are not being 
altered by adoption of the AST, the probability of an accident 
previously evaluated is not affected.
    Plant-specific AST radiological analyses have been performed 
and, based on the results of these analyses, it has been 
demonstrated that the dose consequences of the limiting events 
considered in the analyses are within the acceptance criteria 
provided by the NRC for use with the AST. These criteria are 
presented in 10 CFR 50.67 and Regulatory Guide 1.183. Even though 
the AST dose limits are not directly comparable to the previously 
specified whole body and thyroid dose guidelines of General Design 
Criterion 19 and 10 CFR 100.11, the results of the AST analyses have 
demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore, 
it is concluded that adoption of the AST does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Based on the above discussion, it is concluded that the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Implementation of AST and the proposed changes do not alter or 
involve any design basis accident initiators. These changes do not 
involve any physical changes to the plant and do not affect the 
design function or mode of operations of systems, structures, or 
components in the facility prior to a postulated accident. Since 
systems, structures, and components are operated essentially no 
differently after the AST implementation, no new failure modes are 
created by this proposed change.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes proposed are associated with a new licensing basis 
for analysis of NMP1 DBAs. Approval of the licensing basis change 
from the original source term to the AST is being requested. The 
results of the accident analyses performed in support of the 
proposed changes are subject to revised acceptance criteria. The 
limiting DBAs have been analyzed using conservative methodologies, 
in accordance with the guidance contained in Regulatory Guide 1.183, 
to ensure that analyzed events are bounding and that safety margin 
has not been reduced. The dose consequences of these limiting events 
are within the acceptance criteria presented in 10 CFR 50.67 and 
Regulatory Guide 1.183. Thus, the proposed changes continue to 
ensure that the doses at the exclusion area boundary and low 
population zone boundary, as well as in the control room, are within 
corresponding regulatory criteria.
    Therefore, by meeting the applicable regulatory criteria for 
AST, it is concluded that the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Acting Branch Chief: John P. Boska.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: January 4, 2007.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) 3.7.1, ``Service Water (SW) 
System and Ultimate Heat Sink (UHS),'' as follows: Revise the existing 
Limiting Condition for Operation (LCO) statement to require four 
operable SW pumps to be in operation when SW subsystem supply header 
water temperature is <=82 [deg]F; add a requirement that five operable 
SW pumps be in operation when SW subsystem supply header water 
temperature is >82 [deg]F and <=84 [deg]F; delete Condition G and the 
associated Required Actions and Completion Times; revise Surveillance 
Requirement 3.7.1.3 to increase the maximum allowed SW subsystem supply 
header water temperature from 82 [deg]F to 84 [deg]F; and modify the 
requirements for increasing the surveillance frequency as the 
temperature approaches the limit.
    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.

[[Page 11391]]

    The proposed change eliminates the requirement to perform 
temperature averaging when the UHS temperature is >82 [deg]F, 
establishes 84 [deg]F as the design limit for UHS water temperature 
for operation on a continuous basis, and revises the frequency for 
verifying that the UHS temperature is within the prescribed limit. 
The TS currently allow operation with the UHS water temperature 
temporarily exceeding 82 [deg]F, up to a maximum of 84 [deg]F. The 
UHS temperature itself is not an initiator of accidents analyzed in 
the Updated Safety Analysis Report (USAR). Raising the maximum 
temperature limit and revising the associated surveillance 
requirement frequency do not involve any plant hardware changes or 
new operator actions that could serve to initiate an accident. 
Continuous operation with the elevated UHS temperature may result in 
a few balance-of-plant equipment high temperature alarms. Operator 
response to these alarms would be in accordance with established 
alarm response procedures. In all cases, trip setpoints leading to a 
reactor scram or a power runback will not be reached, and the 
likelihood of component failures that could initiate an accident 
will not be significantly increased.
    The potential impact of the proposed change on the ability of 
the plant to mitigate postulated accidents has been evaluated. These 
evaluations demonstrate that safety-related systems and components 
that rely on the UHS as the cooling medium or as a pump suction 
source are capable of performing their intended safety functions at 
the higher UHS temperature, and that containment integrity and 
equipment qualification are maintained. The calculated post-accident 
dose consequences reflected in the USAR do not directly utilize UHS 
temperature as an input and thus are not impacted by the proposed 
change.
    Based on the above, the proposed change will have no adverse 
effect on plant operation or the availability or operation of any 
accident mitigation equipment. Therefore, the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not alter the current plant 
configuration (no new or different type of equipment will be 
installed) or require any new or unusual operator actions. The 
proposed change will not alter the way any structure, system, or 
component functions and will not cause an adverse effect on plant 
operation or accident mitigation equipment. The response of the 
plant and the operators following a design-basis accident is 
unaffected by the change. The proposed change does not introduce any 
credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases. 
Analyses have shown that the design basis heat removal capability of 
the affected safety-related components is maintained at the 
increased UHS water temperature limit.
    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 margin of safety is determined by the design and 
qualification of the plant equipment, the operation of the plant 
within analyzed limits, and the point at which protective or 
mitigative actions are initiated. The proposed change does not 
impact these factors. An evaluation of the safety systems has been 
performed to ensure their safety functions can be met for operation 
with a UHS water temperature of 84 [deg]F on a continuous basis. 
Operation with the UHS water temperature temporarily exceeding 82 
[deg]F, up to a maximum of 84 [deg]F, is currently allowed. 
Operating on a continuous basis at the higher UHS temperature 
represents a slight reduction in design margins in terms of the 
ability of affected systems to remove accident heat loads. However, 
the evaluation has demonstrated that the proposed change does not 
have a significant impact on the capability of the affected systems 
to perform their safety-related post-accident functions and to 
mitigate accident consequences. The design limits for the 
containment and fuel cladding will not be exceeded, and equipment 
qualification will be maintained. No protection setpoints are 
affected by the proposed change. The revised frequency for 
performing the TS surveillance to verify that the UHS temperature is 
within the prescribed limit will continue to assure that plant 
operators are aware of and are monitoring increasing UHS temperature 
trends prior to reaching a value of 82 [deg]F, when a fifth SW pump 
must be placed in operation. This action is no different than that 
required by the current TS.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Acting Branch Chief: John P. Boska.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: January 29, 2007.
    Description of amendment request: The proposed amendment would 
revise Table 3.3.5.1-1, ``Emergency Core cooling System 
Instrumentation,'' of the Technical Specifications (TSs) to extend the 
quarterly surveillance interval from quarterly to a nominal 24-month 
interval for three low pressure coolant injection loop select logic 
functions. Consistent with the extended test interval, the licensee 
also proposed to change the allowable values associated with each of 
the three logic functions (i.e., response time in seconds). The 
licensee stated that the quarterly surveillance requirement was 
inappropriately introduced when the TSs was converted from its previous 
custom format to the current Improved Technical Specification format by 
Amendment No. 146. Before the conversion, there was no such quarterly 
surveillance requirement. Furthermore, the plant was not designed to 
have these three logic functions tested while on-line.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC). 
The NRC staff reviewed the licensee's analysis, and has performed its 
own analysis as follows:
    (1) Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment would extend the performance interval 
from quarterly to a 24-month interval, and change the associated 
allowable values for the three logic functions. The performance of 
these surveillances, or the failure to perform, as well as the 
surveillance finding (i.e., response time in seconds) are not 
precursors to, and do not affect the probability of, an accident. There 
is no design or operation change associated with the proposed 
amendment. Therefore, the proposed amendment does not increase the 
probability of an accident previously evaluated.
    A delay in performing these surveillances would not result in a 
system being unable to perform its required function. The extended 
surveillance and associated changed allowable values will not affect 
the three logic functions to operate as designed. Therefore, the plant 
systems required to mitigate accidents will remain capable of 
performing their design function. As a result, the proposed amendment 
will not lead to any significant change in the consequences of any 
accident.
    (2) Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 11392]]

    No. The proposed amendment does not involve a physical alteration 
of any system, structure, or component