Biweekly Notice, 67391-67403 [E6-19434]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 224 (Tuesday, November 21, 2006)]
[Notices]
[Pages 67391-67403]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-19434]


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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 October 27, 2006, to November 8, 2006. The 
last biweekly notice was published on November 7, 2006 (71 FR 65139).

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

[[Page 67392]]

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.
    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.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station (Oyster Creek), Ocean County, New Jersey

    Date of amendment request: September 28, 2006.
    Description of amendment request: The amendment would revise the 
Oyster Creek Technical Specifications definition of Channel 
Calibration, Channel Check, and Channel Functional Test in accordance 
with the NUREG-1433, Revision 3, ``Standard Technical Specifications, 
General Electric Plants--BWR [boiling water reactor]/4.''
    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. Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The definitions of Channel Check, Channel Calibration[,] and 
Channel Functional Test specified in Technical Specifications (TS) 
provide basic information regarding what the test involves, the 
components involved in the test, and general information regarding 
how the test is to be performed. Instrument channel checking, 
calibrating, and testing are not initiators of any accident 
previously evaluated. Furthermore, the proposed changes will not 
affect the ability of the channel being checked, calibrated[,] or 
tested to respond as assumed in any accident previously evaluated. 
Therefore, these revised definitions result in no increase in the 
probability of an accident previously evaluated.
    The proposed revisions of these definitions, corresponding 
administrative changes (capitalization of definitions), and the 
proposed alternate testing and calibrating methodology using 
sequential, overlapping testing, and/or actual channel input signals 
and/or in place qualitative assessments of resistance temperature 
detectors (RTD's) and thermocouples (TC's) involve no changes to 
plant design, equipment, or operation related to mitigation of 
accidents. The qualitative evaluation of sensor behavior for non-
adjustable sensors will provide an accurate indication of sensor 
operation and will

[[Page 67393]]

assure that [the evaluated] portion of the channel is operating 
properly, ensuring that the consequences of an accident will remain 
as previously evaluated. Therefore, these revised definitions result 
in no increase in the consequences of an accident previously 
identified.
    Based on the above, AmerGen concludes that the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance of the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed revisions of the instrument surveillance 
definitions, corresponding administrative changes (capitalization of 
definitions), and the proposed alternate testing and calibrating 
methodology using sequential, overlapping testing, and/or actual 
channel input signals and/or in place qualitative assessments of 
RTD's and TC's do not involve a physical alteration of the plant or 
a change in the methods governing normal plant operation. No new or 
different type[s] of equipment will be installed. The proposed 
changes also do not adversely affect the operation or operability of 
existing plant equipment. The proposed revisions will allow a change 
in testing and calibrating methodology. Allowing an alternate 
testing and calibrating methodology will not change how the plant is 
operated. Each instrument channel will be tested one sub channel at 
a time, as is currently performed, and will not create the 
possibility of a new or different kind of accident.
    Based on the above discussion, AmerGen concludes that the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The affected definitions involve checking, calibrating[,] and 
testing of instrumentation used in the mitigation of accidents to 
ensure that the instrumentation will perform as assumed in safety 
analyses. The proposed revisions of these definitions, corresponding 
administrative changes (capitalization of definitions), and the 
proposed alternate testing and calibrating methodology using 
sequential, overlapping testing, and/or actual channel input signals 
and/or in place qualitative assessments of RTD's and TC's does not 
alter the ability of the instrument channel to respond as designed 
or assumed in the safety analyses. As a result[,] the ability of the 
plant to respond to[,] and mitigate[,] accidents is unchanged by the 
revised definitions. 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: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: June 16, 2006, as supplemented by letter 
dated September 14, 2006.
    Description of amendment request: The proposed amendment would 
revise the Byron Station Updated Final Safety Analysis Report (UFSAR) 
to incorporate changes concerning the requirements for physical 
protection from tornado-generated missiles (TGM) for safety-related and 
non-safety related systems and components.
    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 occurrence of the design basis tornado 
remains the same as originally established in the Byron Station 
Updated Final Safety Analysis Report (UFSAR). The request involves 
the use of a probability-based assessment of the need for physical 
tornado missile protection of specific existing features at Byron 
Station.
    The request is to utilize an NRC approved methodology (i.e., the 
Electric Power Research Institute (EPRI) Topical Report ``Tornado 
Missile Risk Evaluation Methodology'') to conclude that the 
acceptance criteria of NUREG-0800, ``Standard Review Plan,'' (SRP) 
Section 2.2.3, ``Evaluation of Potential Accidents,'' Revision 2, 
July 1981, has been met for Byron Station and that tornado missile 
damage of selected components at Byron Station need not be 
considered as a credible event.
    Per Item 2 in Section III of SRP 3.5.1.4, probability methods 
can be used to accept tornado missile effects provided damage to all 
important structures, systems and components, as discussed in 
Regulatory Guide 1.117 are considered. Per Section II of the SRP, 
the acceptance criterion of SRP 2.2.3 is applicable. Section II of 
SRP 2.2.3 states that the expected rate of occurrence of potential 
exposure in excess of 10 CFR Part 100, ``Reactor Site Criteria,'' 
guidelines of approximately 1.0E-06 per reactor year is acceptable, 
if when combined with reasonable qualitative arguments, that the 
realistic probability can be shown to be smaller.
    [The licensee in its September 14, 2006, letter stated the 
following in regards to the consequences of an accident previously 
evaluated:
    The acceptance criteria for the TORMIS analysis has been 
established as 1.0 E-06 per year cumulative probability of a TGM 
striking/damaging an unprotected essential SSC [system, structure or 
component] required for safe shutdown in the event of a tornado, 
which is the same value found to be acceptable by the NRC based on 
the accepted rates of occurrence of potential exposures in excess of 
10 CFR 100 guidelines. This criteria in combination with 
conservative qualitative assumptions show that the realistic 
probability of a potential exposure in excess of the 10 CFR Part 100 
guidelines is lower than 1.0 E-06 per year. The conservative 
qualitative assumptions are the same as previously found to be 
acceptable by the NRC as described below:
    It is assumed that an essential SSC being struck/damaged by a 
tornado missile will result in damage sufficient to preclude it from 
performing its safety function.
    It is assumed that the damage to the essential SSC results in 
damage to fuel sufficient to result in conservatively calculated 
radiological release values in excess of 10 CFR 100 guidelines.
    There are no missiles that can directly impact irradiated fuel, 
even the spent fuel stored in the Spent Fuel Pool.]
    The proposed change is not considered to constitute a 
significant increase in the probability or occurrence or the 
consequences of an accident due to the extremely low probability of 
damage due to tornado-generated missiles and therefore an extremely 
low probability of a radiological release. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of previously evaluated accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change involves the use of an alternative methodology to 
assess the need for tornado missile protection on selected Byron 
Station components. The use of this methodology and the changes to 
the Byron Station UFSAR will be limited to design basis tornado 
applications and do not contribute to the possibility of a new or 
different kind of accident from those previously analyzed.
    No new or different system interactions are created and no new 
processes are introduced. The proposed change does not introduce any 
new failure mechanisms, malfunctions, or accident initiators not 
already considered in the design and licensing bases. Based on this 
evaluation, 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.

[[Page 67394]]

    The changes, allowing for no additional physical protection for 
tornado-generated missiles for certain Byron Sation components, is 
based on successfully meeting the acceptance criteria of NUREG-0800, 
``Standard Review Plan,'' (SRP) Section 2.2.3, ``Evaluation of 
Potential Accidents,'' Revision 2, July 1981. Because of the 
extremely low probability of damage to these components from 
tornado-generated missiles, the change is not considered to 
constitute a significant decrease in the 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, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

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

    Date of amendment request: October 13, 2006.
    Description of amendment request: The proposed amendment would 
eliminate License Condition 2.F, which requires reporting violations of 
Operating License Section 2.C, and eliminates Technical Specification 
5.6.6, which contains a reporting condition similar to Operating 
License Section 2.C.(6).
    The availability of this operating license improvement was 
announced in the Federal Register on November 4, 2005 (70 FR 67202), as 
part of the consolidated line item improvement process (CLIIP). The NRC 
staff issued a notice of opportunity for comment in the Federal 
Register on August 29, 2005 (70 FR 51098), on possible amendments 
concerning this CLIIP, including a model safety evaluation and a 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 November 4, 2005 (70 FR 67202). In its application dated October 13, 
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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the 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 that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the 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 deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a 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: Daniel S. Collins.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: October 5, 2006.
    Description of amendment request: The proposed amendment to the 
Improved Technical Specification will revise the defined pool burnup-
enrichment requirements, storage configuration for fresh fuel and low 
burnup/high enriched fuel, the definition of a peripheral assembly, and 
will include minor editorial changes.
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The LAR proposes to revise the fresh fuel loading configuration. 
PEF [Progress Energy Florida, Inc.] has reanalyzed the criticality 
of the revised storage configuration for fresh fuel checkerboarded 
with spent fuel in Pool A, and surrounded by empty water cells in 
Pool B. Similarly, storage of spent fuel in peripheral storage 
locations, given the new definition, was also reanalyzed. The 
revised fuel storage configuration does not affect any structure, 
system, component or process related to the operation of Crystal 
River Unit 3 (CR-3). As a result, the proposed LAR will not change 
the probability or consequences of any accidents previously 
evaluated that are related to operation of the plant. Thus, only 
those accidents that are related to movement and storage of fuel 
assemblies could be potentially affected by the proposed LAR.
    Fuel Handling Accidents (FHAs) are analyzed in Section 14.2.2.3 
of the CR-3 Final Safety Analysis Report (FSAR). These include a FHA 
inside the Reactor Building (RB) and outside the RB. This LAR 
involves storage of fuel assemblies, an activity conducted outside 
the RB only. Therefore, only the FHA outside the RB event needs to 
be considered.
    The FHA outside the RB event is described as the dropping of a 
fuel assembly into the spent fuel storage pool that results in 
damage to a fuel assembly and the release of the gaseous fission 
products. The current FHA assumes all 208 fuel pins in the dropped 
assembly are damaged and the gas gap activity released. The results 
of that analysis demonstrate that the applicable dose acceptance 
criteria, 10 CFR 50.67 and Regulatory Guide 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors,'' are satisfied. Thus, the consequences of a 
FHA are not increased by the allowed change in the fresh fuel 
configuration. The fresh fuel storage configurations permit more 
effective use of already existing storage locations. They do not 
change the frequency or method for handling fuel assemblies. Fuel 
handling equipment is unaffected. As such, the probability of a FHA 
has not increased. Since only one fuel assembly is handled at a 
time, the consequences of a FHA have not increased.
    The current limiting heat load for the spent fuel pool is from 
the combined impact of stored spent fuel and a full core off-load. 
These changes do not increase spent fuel storage capacity over that 
for which the racks are currently analyzed and it does not increase 
the amount of heat ejected from an off-loaded core. Consequently, 
current analyses for spent fuel pool cooling remain valid. The 
configuration change allows fresh fuel to be checkerboarded with 
spent fuel. Since these changes do not increase the storage capacity 
over that already analyzed for the racks, filling the empty water 
cells in the checkerboard pattern with spent fuel will not increase 
the heat load over that already analyzed. The Pool B allowance to 
surround a higher enriched/lower burnup fuel assembly in Pool B with 
empty water cells or changing the definition of a periphery rack 
cell does not increase the number of spent fuel assembly rack 
locations over that previously analyzed. Therefore, there is no 
increase in the pool heat load over that already analyzed.
    A change in storage configurations in storage Pools A and B does 
not increase the probability of a full core off-load or the 
frequency of establishing maximum heat load conditions.
    The FSAR specifies the normal upper limit of the fuel pool 
cooling system as 160 [deg]F. Administrative controls are 
implemented to

[[Page 67395]]

control when fuel may be moved from the reactor to the fuel pool to 
prevent reaching this limit.
    Because neither the probability nor the consequences of a FHA 
are increased, and because there is not additional heat input to the 
spent fuel pools, it is concluded that the LAR does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Onsite storage of spent fuel assemblies in the spent fuel pools 
is a normal activity for which CR-3 has been designed and licensed. 
As part of assuring that this normal activity can be performed 
without endangering public health and safety, the ability of CR-3 to 
safely accommodate different possible accidents in the spent fuel 
pools, such as dropping a fuel assembly or the misloading of a fuel 
assembly, have been analyzed. The revised fuel storage 
configurations proposed by the LAR does not change the methods of 
fuel movement or fuel storage. No structural or mechanical change to 
racks or fuel handling equipment is being proposed. The proposed 
revisions allow for more effective use of existing, unmodified rack 
locations when fresh or highly enriched, low burnup fuel is stored 
in the pool. The proposed revisions are a modification to the 
criticality analysis only, and therefore the proposed LAR does not 
create any new or different kind of accident from those previously 
evaluated.
    (3) Involve a significant reduction in a margin of safety?
    The CR-3 Improved Technical Specification (ITS) ensures the 
effective neutron multiplication factor, Keff, of the spent fuel 
storage racks is maintained less than or equal to 0.95 when fully 
loaded and flooded with unborated water. The revisions proposed by 
the LAR likewise ensure Keff is maintained less than this 
requirement.
    Analyses for the proposed fuel storage configurations have shown 
that sufficient margin exists for fuel enriched to the maximum 
allowed by the CR-3 license, and for all fuel that is or has been in 
use at CR-3. Maintaining this margin is assured by remaining within 
the limits on initial enrichment and fuel burnup that are specified 
in the CR-3 ITS and, in the case of highly enriched, low burnup fuel 
in Pool B, by water hole spacing. The LAR proposes allowing fresh 
fuel to be checkerboarded with Category B type fuel in Pool A rather 
than with empty water cells. It also allows fresh fuel with high 
initial enrichment which does not meet current burnup requirements 
to be placed in Pool B if surrounded by eight empty water cells. It 
also proposes to change the definition of a periphery rack location 
for storing Category BP type fuel. Analyses show that the new 
proposed limits ensure that Keff remains less than 0.95. Attachment 
E [not included in this FR notice] provides an analysis summary.
    The current CR-3 licensing basis allows the use of 
administrative controls, e.g., curves of initial fuel assembly 
enrichment versus burnup, as a means of preventing criticality in 
the spent fuel pools. The use of these curves would be continued 
under this proposed amendment. The changes to these curves proposed 
by this LAR consist of revising the values of burnup and adding 
notes to restrict loading of certain fuel assemblies to specific 
configurations. These types of curves and administrative controls 
have been included in the CR-3 operating license and their use 
implemented by site procedures for many operating cycles. From this 
previous use, CR-3 personnel are familiar with the practice of using 
administrative controls, such as curves of fuel assembly enrichment 
versus burnup, to prevent criticality events when placing fuel 
assemblies in the spent fuel pool.
    Misloaded and mislocated fuel assemblies were analyzed. The 
analysis demonstrated that misloading of a fresh fuel assembly, 
assuming no soluble poison (boron) in the water does result in 
exceeding the criticality margin regulatory limit of Keff = 0.95. 
The analysis further shows that a concentration of 165 ppm boron in 
the Pool A and a concentration of 46 ppm boron in Pool B is 
sufficient to ensure Keff < 0.95. LCO 3.7.14 currently requires a 
minimum boron concentration of 1925 ppm in the spent fuel pools 
until fuel is verified as having been loaded in accordance with the 
enrichment and burnup requirements of LCO 3.7.15. The soluble boron 
assumed in the analysis for this proposed change is substantially 
less than the 1925 ppm required by the existing license. Therefore, 
existing license requirements for soluble boron remain conservative.
    The NRC staff has reviewed the analysis provided for Florida Power 
Corporation 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 (Acting): L. Raghavan.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center (DAEC), Linn County, Iowa

    Date of amendment request: July 17, 2006.
    Description of amendment request: The proposed amendment would 
revise the Limiting Condition for Operation (LCO) 3.6.3.1 to eliminate 
the requirement for the Containment Atmospheric Dilution (CAD) system, 
allowing its removal from the DAEC. LCO 3.6.3.2 would also be revised 
to allow an additional 48 hours on plant start-up or shutdown sequences 
for the primary containment to be de-inerted.
    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 Containment Atmosphere Dilution (CAD) system and primary 
containment oxygen concentration are not initiators to any accident 
previously evaluated in the DAEC Updated Final Safety Analysis 
Report (UFSAR). The CAD system and containment oxygen concentration 
were previously relied upon to mitigate the consequences of a design 
basis accident (DBA) combustible gas mixture. However, the revised 
10 CFR 50.44 (68 FR 54123) no longer defines a DBA hydrogen release 
(i.e., combustible gas mixture) and the Commission has subsequently 
found that the DBA loss of coolant accident (LOCA) hydrogen release 
is not risk significant. In addition, hydrogen control systems, such 
as CAD, have been determined to be ineffective at mitigating 
hydrogen releases from the more risk significant beyond design basis 
accidents that could threaten containment integrity. Therefore, 
elimination of the CAD system will not significantly increase the 
consequences of any accident previously evaluated. The consequences 
of an accident while relying on the revised Required Actions for 
primary containment oxygen concentration are no different than the 
consequences of the same accidents under the current Required 
Actions. As a result, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant, except for the elimination of the CAD system (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. The CAD system is not 
considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building from any DBA. In addition, the changes do 
not impose any new or different requirements. The changes to the 
Technical Specifications for oxygen concentration do not alter 
assumptions made in the safety analysis, but reflect changes to the 
safety analysis requirements allowed under the revised 10 CFR 50.44. 
Specifically that an inerted containment is no[t] required to 
mitigate any DBA, but has been found to be helpful in mitigating 
certain beyond design basis events (i.e., severe accidents) that 
could generate combustible levels of hydrogen.

[[Page 67396]]

    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The installation of combustible gas control systems, such as 
CAD, required by the original Sec.  50.44(b)(3) was intended to 
address the limited quantity and rate of hydrogen generation that 
was postulated from a design-basis LOCA. The Commission has found 
that this hydrogen release is not risk-significant because the 
design-basis LOCA hydrogen release does not contribute to the 
conditional probability of a large release up to approximately 24 
hours after the onset of core damage. In addition, these systems 
were ineffective at mitigating hydrogen releases from risk-
significant accident sequences that could threaten containment 
integrity. (68 FR 54123). The proposed changes to CAD and primary 
containment oxygen concentration reflect this new regulatory 
position and, in light of the remaining plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, including 
postulated beyond design basis events, does not result in a 
significant reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. R. E. Helfrich, Florida Power & Light 
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: September 15, 2006.
    Description of amendment request: The proposed amendment would 
replace the current control system and it will increase the nominal 
control fluid oil operating pressure from 114 pounds per square inch 
gauge (psig) to 1600 psig. The control fluid oil pressure provides an 
input to the reactor protection system via three pressure switches 
connected to the control fluid header. Due to the change in the 
operating pressure, I&M is proposing a revision to the allowable low 
fluid oil pressure value from greater than or equal to 57 psig to 
greater than or equal to 750 psig.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change reflects a design change to the turbine 
control system that increases the control oil pressure, 
necessitating a change to the value at which a low fluid oil 
pressure initiates a reactor trip. The turbine control oil pressure 
is an input to the reactor trip instrumentation, and the reactor 
trip is a response to an event that trips the turbine. A change in 
the nominal control oil pressure does not introduce any mechanisms 
that would increase the probability of an accident previously 
analyzed. The reactor trip on turbine trip function is an 
anticipatory trip, and the safety analysis does not credit this trip 
for protecting the reactor core. Thus, the consequences of 
previously analyzed accidents are not impacted.
    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 control fluid oil pressure decreases in response to a 
turbine trip. The value at which the low control fluid oil initiates 
a reactor trip is not an accident initiator. The change in the value 
reflects the higher pressure of the turbine control system that will 
be installed during the Unit 2 Cycle 17 refueling outage.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The change involves a parameter that initiates an anticipatory 
reactor trip following a turbine trip. The safety analyses do not 
credit this anticipatory trip for reactor core protection.
    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 this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment requests involve no 
significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Acting Branch Chief: Martin C. Murphy.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit 1 and 2, Berrien County, Michigan

    Date of amendment request: September 15, 2006.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) to change Required Action 
Notes in TS 3.3.1, ``Reactor Trip System Instrumentation,'' and TS 
3.3.2, ``Engineered Safety Features Actuation System Instrumentation,'' 
to reflect installed bypass test capability, as well as correct one 
administrative error in TS 3.3.1 Condition Q. The proposed changes to 
the Required Action Notes are consistent with wording in Standard 
Technical Specifications (NUREG-1431, Revision 3) for plants with 
installed bypass test capability.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change reflects NUREG-1431, Revision 3, ``Standard 
Technical Specifications, Westinghouse Plants,'' (STS) wording for 
plants with installed bypass test capability and aligns Technical 
Specification (TS) Condition entry requirements with other portions 
of the TS. The proposed changes do not modify how the reactor trip 
system (RTS) and engineered safety features actuation systems 
(ESFAS) functions respond to an accident condition. The proposed 
changes to the TS Required Action Notes prevent unnecessary TS 
Action entry during performance of surveillance testing. The 
probability of accidents previously evaluated remains unchanged 
since the proposed change does not affect any accident initiators. 
The consequences of accidents previously evaluated are unaffected by 
this change because no change to any accident mitigation scenario 
has resulted and there are no additional challenges to fission 
product barrier integrity.
    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.
    No changes are being made to the plant that would introduce any 
new accident causal mechanisms. The proposed change to

[[Page 67397]]

the Required Action Notes and Condition entry requirements does not 
adversely affect previously identified accident initiators and does 
not create any new accident initiators. The change does not affect 
how the RTS and ESFAS functions operate. No new single failure or 
accident scenarios are created by the proposed change and the 
proposed change does not result in any event previously deemed 
incredible being made credible.
    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.
    No safety analyses were changed or modified as a result of the 
proposed TS changes to reflect STS wording for plants with installed 
bypass test capability or for aligning TS Condition entry 
requirements. All margins associated with the current safety 
analyses acceptance criteria are unaffected. The current safety 
analyses remain bounding. The safety systems credited in the safety 
analyses will continue to be available to perform their mitigation 
functions. The proposed change does not affect the availability or 
operability of safety-related systems and components.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: August 14, 2006.
    Description of amendment request: The proposed amendments would 
make miscellaneous improvements to the Technical Specifications (TS) 
for Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2. The 
proposed amendments would revise TS 1.3, ``Completion Times''; TS 
3.1.4, ``Rod Group Alignment Limits''; TS 3.3.7, ``Spent Fuel Pool 
Special Ventilation System (SFPSVS) Actuation Instrumentation''; TS 
3.7.10, ``Control Room Special Ventilation System (CRSVS)''; and TS 
Chapter 4.0, ``Design Features''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes changes to the Prairie 
Island Nuclear Generating Plant Technical Specifications as follows: 
Technical Specification 1.3, ``Completion Times'', revise a text 
header and add a new text header; Technical Specification 3.1.4, 
``Rod Group Alignment Limits'', remove a Surveillance Note which 
cross-references another Technical Specification and may cause 
confusion; Technical Specification 3.3.7, ``Spent Fuel Pool Special 
Ventilation System (SFPSVS) Actuation Instrumentation'', revises the 
Modes of Applicability consistent with plant design and the 
Technical Specifications for the Spent Fuel Pool Special Ventilation 
System, the supported system; Technical Specification 3.7.10, 
``Control Room Special Ventilation System (CRSVS)'', revises the 
applicability of Condition C and clarifies the requirements of the 
Surveillance to verify train filtration flow; and Technical 
Specification Chapter 4.0, ``Design Features'', revises Reference 1 
to the most recent version of the document.
    Revising and adding text headers in Technical Specification 1.3 
are administrative changes because the revised document does not 
change any basis for the current Technical Specifications. Since 
these are administrative changes, they do not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident. Technical Specification 3.1.4 assures that the 
control rod positions are within the limits assumed in the safety 
analysis and that the assumed shutdown margin is available when 
needed. This license amendment request proposes to remove a Note 
from a surveillance requirement that cross-references to Technical 
Specification 3.1.7. Removal of this Note does not change plant 
operations, testing or maintenance; therefore the proposed change 
does not involve a significant increase in the probability of an 
accident. Since plant operations, testing and maintenance are not 
changed, the proposed changes do not involve a significant increase 
in the consequences of an accident previously evaluated.
    The Spent Fuel Pool Special Ventilation System filters 
radioactive materials in the fuel pool enclosure atmosphere released 
following a fuel handling accident. This license amendment request 
proposes to revise the Modes and Other Specified Conditions of 
Applicability for the actuation instrumentation.
    Technical Specification to be consistent with the Modes and 
Other Specified Conditions of Applicability in the Technical 
Specification for the supported system. The Spent Fuel Pool Special 
Ventilation System and its actuation instrumentation are not 
accident initiators; therefore, the proposed changes do not affect 
the probability of an accident. With the proposed change, the 
Technical Specifications will continue to require the system 
actuation instrumentation to be operable when irradiated fuel is 
moved in the fuel pool enclosure which is also the required 
Applicability in the supported system Technical Specification. Since 
the instrumentation will be required to actuate the supported system 
when it is required to operate, the accident consequences will 
continue to be mitigated with this proposed Technical Specification 
change. Thus, the proposed Technical Specification change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    The Control Room Special Ventilation System provides an enclosed 
control room environment from which the plant can be operated 
following an uncontrolled release of radioactivity. This system is 
not an accident initiator, thus the proposed changes do not increase 
the probability of an accident. This license amendment proposes 
changes which will: (1) Reduce the time to shut down the plant when 
Technical Specification required actions or completion time is not 
met; and (2) clarifies surveillance requirements to assure that the 
system performs as designed. These changes do not impact the 
performance of the system; thus this change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Updating the reference in Technical Specification Chapter 4.0 is 
an administrative change because the revised document does not 
change any basis for the current Technical Specifications. Since 
this is an administrative change, it does not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    The changes proposed in this license amendment do not involve a 
significant increase the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment request proposes changes to the Prairie 
Island Nuclear Generating Plant Technical Specifications as follows: 
Technical Specification 1.3, ``Completion Times'', revise a text 
header and add a new text header; Technical Specification 3.1.4, 
``Rod Group Alignment Limits'', remove a Surveillance Note which 
cross-references another Technical Specification and may cause 
confusion; Technical Specification 3.3.7, ``Spent Fuel Pool Special 
Ventilation System (SFPSVS) Actuation Instrumentation'', revises the 
Modes of Applicability consistent with plant design and the 
Technical Specifications for the Spent Fuel Pool Special Ventilation 
System, the supported system; Technical Specification 3.7.10, 
``Control Room Special Ventilation System (CRSVS)'', revises the 
applicability of Condition C and clarifies the requirements of the 
Surveillance to verify train filtration flow; and Technical 
Specification Chapter 4.0, ``Design Features'',

[[Page 67398]]

revises Reference 1 to the most recent version of the document.
    Revising and adding text headers in Technical Specification 1.3 
are administrative changes because the revised document does not 
change any basis for the current Technical Specifications. Since 
these are administrative changes, they do not create the possibility 
of a new or different kind of accident.
    Removal of a surveillance note from Technical Specification 
3.1.4 that cross-references another Technical Specification does not 
change any plant operations, maintenance activities or testing 
requirements. The Limiting Conditions for Operation will continue to 
be met and the proper control rod positions will continue to be 
maintained. There are no new failure modes or mechanisms created 
through the removal of the Surveillance Requirements Note, nor are 
new accident precursors generated by this change. This proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed revision of Modes of Applicability for the Spent 
Fuel Pool Special Ventilation System actuation instrumentation makes 
operation of the actuation instrumentation consistent with the 
Technical Specification requirements for the supported system and 
does not change the operation of the supported system for accident 
mitigation. The Limiting Conditions for Operation will continue to 
be met, no new failure modes or mechanisms are created and no new 
accident precursors are generated by this change. This proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The changes proposed for the Control Room Special Ventilation 
System Technical Specifications do not change any the system 
operations, maintenance activities or testing requirements. The 
Limiting Conditions for Operation will continue to be met, no new 
failure modes or mechanisms are created and no new accident 
precursors are generated by this change. This proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Updating the reference in Technical Specification Chapter 4.0 is 
an administrative change because the revised document does not 
change any basis for the current Technical Specifications. Since 
this is an administrative change, it does not create the possibility 
of a new or different kind of accident.
    The Technical Specification changes proposed in this license 
amendment do not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment request proposes changes to the Prairie 
Island Nuclear Generating Plant Technical Specifications as follows: 
Technical Specification 1.3, ``Completion Times'', revise a text 
header and add a new text header; Technical Specification 3.1.4, 
``Rod Group Alignment Limits'', remove a Surveillance Note which 
cross-references another Technical Specification and may cause 
confusion; Technical Specification 3.3.7, ``Spent Fuel Pool Special 
Ventilation System (SFPSVS) Actuation Instrumentation'', revises the 
Modes of Applicability consistent with plant design and the 
Technical Specifications for the Spent Fuel Pool Special Ventilation 
System, the supported system; Technical Specification 3.7.10, 
``Control Room Special Ventilation System (CRSVS)'', revises the 
applicability of Condition C and clarifies the requirements of the 
Surveillance to verify train filtration flow; and Technical 
Specification Chapter 4.0, ``Design Features'', revises Reference 1 
to the most recent version of the document.
    Revising and adding text headers in Technical Specification 1.3 
are administrative changes because the revised document does not 
change any basis for the current Technical Specifications. Since 
these are administrative changes, they do not involve a significant 
reduction in a margin of safety.
    Plant operations are required to meet all Technical 
Specifications for which the Applicability is met; therefore, 
removal of the cross-reference Note from a Technical Specification 
3.1.4 surveillance requirement does not change how the plant is 
operated and therefore, this change does not involve a significant 
reduction in a margin of safety.
    Technical Specification 3.3.7 provides requirements for 
actuation instrument which supports the operation of the Spent Fuel 
Pool Special Ventilation System as required by Technical 
Specification 3.7.13. The current Applicability for Technical 
Specification 3.3.7 requires the actuation instrumentation to be 
operable in Modes which are not required by Technical Specification 
3.7.13. This license amendment proposes to make Technical 
Specification 3.3.7 Applicability the same as Technical 
Specification 3.7.13. This change does not reduce the conditions or 
Modes when the Spent Fuel Pool Special Ventilation System will 
operate and perform its accident mitigation function; thus this 
change does not involve a significant reduction in a margin of 
safety.
    This license amendment proposes changes to the Control Room 
Special Ventilation System Technical Specifications which will: (1) 
Reduce the time to shut down the plant when Technical Specification 
required actions or completion time is not met; and (2) clarifies 
surveillance requirements to assure that the system performs as 
designed. The proposed time to shut down the plant is consistent 
with other Technical Specifications for shutting down the plant and 
allows adequate time for an orderly shut down of the plant; thus 
this change does not involve a significant reduction in a margin of 
safety. The surveillance requirement clarifications do not reduce 
any testing requirements and will continue to demonstrate that the 
system can perform its required safety function and satisfy the 
Limiting Conditions for Operation. Thus this change does not involve 
a significant reduction in a margin of safety.
    Updating the reference in Technical Specification Chapter 4.0 is 
an administrative change because the revised document does not 
change any basis for the current Technical Specifications. Since 
this is an administrative change, it does not involve a significant 
reduction in a margin of safety.
    The Technical Specification changes proposed in this license 
amendment 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 requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: M. Murphy (A).

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama

    Date of amendment request: October 26, 2006.
    Description of amendment request: The proposed request would revise 
the Units 2 and 3 emergency diesel generator (EDG) Technical 
Specification (TS) Completion Time (CT) from 14 days to 7 days for 
restoration of an inoperable EDG. The current 14-day CT was based on 
the assumption that Unit 1 was shut down. The near-term restart of Unit 
1 will invalidate this assumption, therefore, the affected CTs are 
being returned to their original duration of 7 days.
    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 Technical Specification change involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    Response: No.
    The EDGs are designed as backup alternating current (AC) power 
sources in the event of a loss of offsite power. The proposed 
restoration of the EDG CT to its original TS duration does not 
change the conditions, operating configurations, or minimum amount 
of operating equipment assumed in the safety analysis for accident 
mitigation. No changes are proposed in the manner in which the EDGs 
provide plant protection or which create new modes of plant 
operation. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences