Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 26173-26184 [E7-8679]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 88 (Tuesday, May 8, 2007)]
[Notices]
[Pages 26173-26184]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-8679]


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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 April 13, 2007 to April 26, 2007. The last 
biweekly notice was published on April 24, 2007 (72 FR 20375).

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

[[Page 26174]]

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: November 27, 2006.
    Description of amendment request: The amendment would revise the 
Oyster Creek Technical Specification (TS) 6.9.1.d, ``Annual Radioactive 
Effluent Release Report,'' by changing the requirement to submit the 
report within 60 days of January 1. Specifically, the revised 
requirement would be to submit the report prior to May 1 of each year.
    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 involves a revision to the required 
submittal date for the Radioactive Effluent Release Report, and is 
administrative in nature. The change will not alter the physical 
design or operation of any plant structure, system, or component.
    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?
    Response: No.
    The proposed change is administrative in nature. The proposed 
change has no impact on the design, function or operation of any 
plant structure, system or component and does not affect any 
accident analyses. Accordingly, the change does not introduce any 
new accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure, system, or component to perform 
their safety function.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change is administrative in nature, does not negate 
any existing requirement, and does not adversely affect existing 
plant safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there is no change being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant 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: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 26175]]

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: March 28, 2007.
    Description of amendment request: The proposed change would revise 
the required wattage specified in the River Bend Station, Unit 1 (RBS), 
Technical Specification 5.5.7.e, Ventilation Filter Testing Program, 
for the Control Room Fresh Air System (CRFAS) heater for testing. The 
proposed required wattage for testing the CRFAS heater would be revised 
from 23  2.3 kilowatt (kW), to ``>==15 kW.''
    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.
    This change specifies the required power (in kW) for the Control 
Room ventilation electric heaters to decrease relative humidity of 
the air to less than 70% relative humidity as required for proper 
operation of the charcoal absorber components based on calculated 
requirements. The heater will continue to perform its intended 
design function as designed. The heater is not an accident 
precursor.
    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 heater will continue to perform its function as designed. 
The heater provides humidity control for the Control Room filter 
unit during a design basis accident. Changing the test acceptance 
criteria to a calculated value has no influence on, nor does it 
contribute in any way to, the possibility of a new or different kind 
of accident or malfunction from those previously analyzed. No change 
has been made to the design, function or method of performing 
testing. No safety-related equipment or safety functions are altered 
as a result of this change.
    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 margin of safety is changed as a result of this change. The 
heater will continue to perform its design function. Testing 
methodology has not changed. The function of the heater is 
unchanged. The acceptance criterion has been changed to a calculated 
value rather than the name plate rating to make testing more 
realistic. The heater will continue to operate to perform its 
intended design function.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: March 30, 2007.
    Description of amendment request: The proposed amendment would 
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification 
(TS) to support a partial re-rack of the storage racks in the ANO-2 
spent fuel pool (SFP). The proposed amendment would revise TS 3.9.12, 
``Fuel Storage,'' and its associated tables, figures, and surveillance 
requirements, TS 5.3, ``Fuel Storage,'' and add TS 6.5.17, ``Metamic 
Coupon Sampling Program.'' The ANO-2 TS 3.9.12 would be changed to: (1) 
Support higher fuel assembly U-235 enrichment; (2) apply the 
appropriate loading restrictions; and (3) delete the dry cask loading 
restrictions. ANO-2 TS 5.3.1b would be changed to reflect a different 
SFP boron concentration that is needed to assure K-effective 
(Keff) remains less than or equal to 0.95. ANO-2 TS 5.3.2a 
would be modified to reflect a higher fuel assembly U-235 enrichment. A 
new coupon sampling program would be added as TS 6.5.17. In addition, 
TS Surveillance Requirement 4.9.12.d would be added to direct 
performance of the coupon sampling program.
    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.
    Fuel Handling Accidents
    The current licensing bases for the dose consequences associated 
with a fuel handling accident (FHA), which was performed considering 
a maximum U-235 enrichment of 5.0 wt% and a maximum burnup of 65 
megawatt-days/kilograms of uranium, does not exceed 25% of 10 CFR 
[Title 10 of the Code of Federal Regulations] 100 limits. The 
proposed change is bounded by the current analysis and therefore, 
there is no increase in the dose consequences associated with a[n] 
FHA.
    During rack removal and installation, safe load paths will be 
determined and written procedures followed to ensure that the racks 
are not carried over any fuel assemblies. With the proposed 
limitations on rack and cask movement, there should be no impact to 
spent fuel and no radiological consequences due to fuel rack 
installation. The racks will be moved with a single failure proof 
crane. Therefore, a postulated drop of a rack is not a credible 
accident.
    The probability of having a[n] FHA has not increased.

Criticality Accidents Associated With a Dropped Fuel Assembly

    The three fuel assembly drop accidents described below can be 
postulated to increase reactivity. However, for these accident 
conditions, the double contingency principle of ANS [American 
National Standard] N16.1-1975 is applied. This states that it is 
unnecessary to assume two unlikely, independent, concurrent events 
to ensure protection against a criticality accident. Thus, for 
accident conditions, the presence of soluble boron in the storage 
pool water can be assumed as a realistic initial condition since its 
absence would be a second unlikely event.
    Three types of drop accidents have been considered: a vertical 
drop accident, a horizontal drop accident, and an inadvertent drop 
of an assembly between the outside periphery of the rack and the 
pool wall. The structural damage to the pool liner, the racks, and 
fuel assembly resulting from a dropped fuel assembly striking the 
rack, the pool floor, or another assembly located in the racks is 
primarily dependent on the mass of the falling object, drop height, 
and structural configuration of the rack. The two parameters related 
to the fuel assembly (mass and drop height) are not changed by the 
proposed rack modification. The new rack design was evaluated for 
all postulated structural drops and the structural damage to these 
items remains within acceptable limits. In all cases the proposed TS 
limit for boron concentration ensures that a five percent 
subcriticality margin is met for the postulated accidents.

Criticality Accidents Associated With a Misplaced Fuel Assembly

    The fuel assembly misplacement accident was considered for all 
storage configurations. An assembly with high reactivity is assumed 
to be placed in a storage location which requires restricted storage 
based on initial U-235 loading, cooling time, and burnup. The 
presence of boron in the pool water assumed in the analysis has been 
shown to offset the worst case reactivity effect of a misplaced

[[Page 26176]]

fuel assembly for any configuration. This boron requirement is less 
than the boron concentration required by the ANO-2 TS. Thus, a five 
percent subcriticality margin is met for postulated accidents, since 
any reactivity increase will be much less than the negative worth of 
the dissolved boron.

Optimum Moderation Accident

    For fuel storage applications in the SFP, water is usually 
present. An ``optimum moderation'' accident is not a concern in SFP 
storage racks because the rack design prevents the preferential 
reduction of water density between the cells of a rack (e.g., 
boiling between cells). In addition, the criticality analysis has 
demonstrated that the effective neutron multiplication factor 
(Keff) will remain less than 1.0 when the SFP is fully 
flooded with unborated water.
    An ``optimum moderation'' accident in the new fuel vault was 
evaluated and the conclusions of that evaluation confirmed that the 
reactivity effect is less than the regulatory limit of 0.98 for 
Keff.

Loss of SFP Cooling

    The proposed modification to the ANO-2 SFP racks does not result 
in a change to the SFP cooling system and therefore the probability 
of a loss of SFP cooling is not increased.
    The consequences of a loss of spent fuel pool cooling were 
evaluated and found to not involve a significant increase as a 
result of the proposed changes. A thermal-hydraulic evaluation for 
the loss of SFP cooling was performed. The analysis determined that 
the minimum time to boil is about two hours following a complete 
core off load and a complete loss of forced cooling. This provides 
sufficient time for the operators to restore cooling or establish an 
alternate means of cooling before the water shielding above the top 
of the racks falls below 10 feet. Therefore, the proposed change 
represents no increase in the consequences of loss of pool cooling.

Seismic Event

    The proposed rack modification does not result in an increase in 
the probability or consequences of a design basis seismic event. The 
new racks were analyzed and all structural acceptance criteria are 
shown to be met during seismic events. The structural capability of 
the SFP and liner will not be exceeded as a result of the new rack 
design.
    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 presence of soluble boron in the pool water assumed in the 
criticality analysis is less than the boron concentration required 
by the ANO-2 TSs. Thus, a five percent subcriticality margin is met 
for postulated accidents, since any reactivity increase will be much 
less than the negative worth of the dissolved boron.
    No new or different types of fuel assembly drop scenarios are 
created by the proposed change. During the installation of the new 
racks, the possibility of dropping a rack is not a credible accident 
since a single failure proof crane and safe load paths will be used 
for rack movements. No new or different fuel assembly misplacement 
accidents will be created. Administrative controls currently exist 
to assist in assuring fuel misplacement does not occur.
    No changes are proposed to the spent fuel pool cooling system or 
makeup systems and therefore no new accidents are considered related 
to the loss of cooling or makeup capability.
    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.
    With the presence of a nominal boron concentration, the SFP 
storage racks will be designed to assure a subcritical array with a 
five percent subcritical margin (95% probability at the 95 % 
confidence level). This has been verified by criticality analyses.
    Credit for soluble boron in the SFP water is permitted under 
accident conditions. The proposed modification that will allow 
installation of the new racks does not result in the potential of 
any new misplacement scenarios. Criticality analyses have been 
performed to determine the required boron concentration that would 
ensure the maximum Keff does not exceed 0.95. The ANO-2 
TS for the minimum SFP boron concentration is greater than that 
required to ensure Keff remains below 0.95. Therefore, 
the margin of safety defined by taking credit for soluble boron will 
be maintained.
    The structural analysis of the new spent fuel racks along with 
the evaluation of the SFP structure indicated that the integrity of 
these structures will be maintained. The structural requirements 
were shown to be satisfied, thus the safety margins were maintained.
    In addition the proposed change includes a coupon sampling 
program that will monitor the physical properties of the 
MetamicTM absorber material. The monitoring program 
provides a method of verifying that the assumptions used in the SFP 
criticality analyses remain valid.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: March 1, 2007.
    Description of amendment request: The proposed change would revise 
Grand Gulf Nuclear Station, Unit 1, Technical Specification (TS) Tables 
3.3.5.1-1 and 3.3.5.2-1 to modify the allowable values of the low 
Condensate Storage Tank (CST) level setpoints for the High Pressure 
Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) suction 
swap from the CST to the Suppression Pool. The change is necessary to 
correct an error in the original plant design. The error, under certain 
conditions, could prevent a swap of the HPCS and RCIC suction flow 
paths to the Suppression Pool. Currently, the erroneous setpoints have 
been corrected to a higher level, and are administratively controlled 
in accordance with the Administrative Letter 98-10, ``Dispositioning of 
Technical Specifications That Are Insufficient To Assure Plant 
Safety.''
    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.
    This change will adjust the setpoint for an automatic swap of 
the suction for the HPCS and RClC systems from the Condensate 
Storage Tank (CST) to the Suppression Pool. The Suppression Pool is 
the source of water credited in the accident analyses. This transfer 
is not the initiator of any analyzed accident. The setpoint 
adjustment will allow a transfer of the suction to an assured 
safety-related water source earlier in the event and will have no 
effect on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Transfer of the suction source for HPCS and RClC will occur 
sooner as a result of this change. No new operational conditions 
beyond those currently allowed are introduced. This change is 
consistent with the safety analyses assumptions and current

[[Page 26177]]

plant operating practices. This simply corrects the setpoint 
consistent with the accident analyses and therefore cannot create 
the possibility of a new or different kind of accident from any 
previously evaluated accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not reduce safety, but rather allows 
the transfer from the CST to the Suppression Pool sooner. The 
Suppression Pool is the source of water credited in the accident 
analyses. This change is consistent with the safety analyses 
assumptions and current plant operating practices. No new 
operational conditions beyond those currently allowed are created by 
these changes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: October 18, 2006.
    Description of amendment request: A change is proposed to the 
technical specifications (TSs) of LaSalle County Station, Units 1 and 2 
(LaSalle), consistent with TS Task Force Traveler No. 432 (TSTF-423), 
``Technical Specification End States, NEDC-32988-A,'' to the standard 
TSs for boiling-water reactor plants, to allow for some systems entry 
into hot shutdown rather than cold shutdown, to repair equipment if 
risk is assessed and managed consistent with the program in place for 
complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.65(a)(4). The proposed amendment would 
modify the TS to risk-informed requirements regarding selected required 
action end states provided in TSTF-423, Revision 0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows a change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1) Those end states where entry 
into the shutdown mode is for a short interval, (2) entry is 
initiated by inoperability of a single train of equipment or a 
restriction on a plant operational parameter, unless otherwise 
stated in the applicable technical specification, and (3) the 
primary purpose is to correct the initiating condition and return to 
power operation as soon as is practical. Risk insights from both the 
qualitative and quantitative risk assessments were used in specific 
TS assessments. Such assessments are documented in Section 6 of GE 
NEDC-32988, Revision 2, ``Technical Justification to Support Risk 
Informed Modification to Selected Required Action End States for BWR 
Plants.'' They [risk assessments] provide an integrated discussion 
of deterministic and probabilistic issues, focusing on specific 
technical specifications, which are used to support the proposed TS 
end state and associated restrictions. The [NRC] staff finds that 
the risk insights support the conclusions of the specific TS 
assessments. Therefore, the probability of an accident previously 
evaluated is not significantly increased, if at all. The 
consequences of an accident after adopting proposed TSTF-423, are no 
different than the consequences of an accident prior to adopting 
TSTF-423. 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.

Criterion 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). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded, i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment, 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 and the commitment by the licensee to adhere to the guidance 
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0, 
``Technical Specifications End States, NEDC-32988-A,'' 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.

Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The [Boiling Water Reactor Owners Group] 
BWROG's risk assessment approach is comprehensive and follows [NRC] 
staff guidance as documented in [Regulatory Guides] RGs 1.174 and 
1.177. In addition, the analyses shows that the criteria of the 
three-tiered approach for allowing TS changes are met. The risk 
impact of the proposed TS changes was assessed following the three-
tiered approach recommended in RG 1.177. A risk assessment was 
performed to justify the proposed TS changes. The net change to the 
margin of safety is insignificant. Therefore, this change does not 
involve a significant reduction in a margin of safety.

    LaSalle has reviewed the proposed no significant hazards 
consideration determination published on March 23, 2006, (71 FR 14743) 
as part of the consolidated line item improvement 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: Russell Gibbs.

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: February 28, 2007.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 5.2.2, ``Plant Staff'', and TS 5.3, 
``Plant Staff Qualifications'', requirements for shift technical 
advisor (STA) qualifications. The proposed changes will specify that 
personnel who perform the function of STA shall meet the qualification 
requirements of the Commission Policy Statement on Engineering 
Expertise on Shift, published in the Federal Register on October 28, 
1985 (50 FR 43621). This change will allow qualified personnel to 
perform the function of STA without

[[Page 26178]]

also holding a senior reactor operator (SRO) license.
    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 to add a new sentence to 
Technical Specification 5.2.2 specifying that personnel who perform 
the function of shift technical advisor shall meet the qualification 
requirements of the Commission Policy Statement on Engineering 
Expertise on Shift and remove shift technical advisor qualification 
requirements from Technical Specification 5.3.1. This change will 
allow qualified personnel to perform the function of shift technical 
advisor without also holding a senior reactor operator license.
    The proposed changes are administrative changes to Technical 
Specifications Chapter 5, the administrative chapter of the 
Technical Specifications. Shift technical advisors perform the 
function of on-shift technical advisor to the shift supervisor and 
do not operate the plant. Therefore, the changes proposed in this 
license amendment request do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment request proposes to add a new sentence to 
Technical Specification 5.2.2 specifying that personnel who perform 
the function of shift technical advisor shall meet the qualification 
requirements of the Commission Policy Statement on Engineering 
Expertise on Shift and remove shift technical advisor qualification 
requirements from Technical Specification 5.3.1. This change will 
allow qualified personnel to perform the function of shift technical 
advisor without also holding a senior reactor operator license.
    The Technical Specification changes proposed in this license 
amendment are administrative, do not change the manner in which the 
plant is operated, and 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 to add a new sentence to 
Technical Specification 5.2.2 specifying that personnel who perform 
the function of shift technical advisor shall meet the qualification 
requirements of the Commission Policy Statement on Engineering 
Expertise on Shift and remove shift technical advisor qualification 
requirements from Technical Specification 5.3.1. This change will 
allow qualified personnel to perform the function of shift technical 
advisor without also holding a senior reactor operator license.
    The proposed changes are administrative changes to Technical 
Specifications Chapter 5, the administrative chapter of the 
Technical Specifications. Shift technical advisors perform the 
function of on-shift technical advisor to the shift supervisor and 
do not operate the plant. Thus, the Technical Specification changes 
proposed in this license amendment request 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: L. Raghavan.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: January 18, 2007.
    Brief description of amendments: The amendments requested would 
revise Technical Specifications (TS) requirement 3.8.1, ``AC Sources--
Operating,'' Extension of Completion Times for Offsite Circuits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification (TS) Completion Time (CT) 
extension does not significantly increase the probability of 
occurrence of a previously evaluated accident because the startup 
transformers (STs) are not initiators of previously evaluated 
accidents involving a loss of offsite power (LOOP). The proposed 
changes to the TS Required Actions CTs do not affect any of the 
assumptions used in the deterministic or the PSA [probabilistic 
safety assessment] analysis relative to LOOP initiating event 
frequency. Implementation of the proposed changes does not result in 
a risk significant impact. The onsite AC [alternating current] power 
sources will remain highly reliable and the proposed changes will 
not result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety as 
measured by the increase in core damage frequency (CDF) is less than 
1E-06 per year and the increase in large early release frequency 
(LERF) is less than 1E-07 per year. In addition, for the CT changes, 
the incremental conditional core damage probabilities (ICCDP) and 
incremental conditional large early release probabilities (ICLERP) 
are less than 5E-07 and 5E-08, respectively. These changes meet the 
acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, 
since the onsite AC power sources will continue to perform their 
functions with high reliability as originally assumed and the 
increase in risk as measured by [Delta]CDF, [Delta]LERF, ICCDP, and 
ICLERP risk metrics is within the acceptance criteria of existing 
regulatory guidance, there will not be a significant increase in the 
consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes are consistent with safety analysis assumptions 
and resultant consequences.
    The proposed TS CT extension will continue to provide assurance 
that the sources of power to 6.9 kV [kilovolts] AC buses perform 
their function when called upon. Extending the TS CT to 30 days does 
not affect the design of the STs, the operational characteristics of 
the STs, the interfaces between the STs and other plant systems, the 
function, or the reliability of the STs. Thus, the STs will be 
capable of performing their accident mitigation functions and there 
is no impact to the radiological consequences of any accident 
analysis.
    The Configuration Risk Management Program (CRMP) in TS 5.5.18 is 
an administrative program that assesses risk based on plant status. 
The risk-informed CT will be implemented consistent with the CRMP 
and approved plant procedures. When utilizing the 30-day extension, 
requirements of the CRMP per TS 5.5.18 call for the consideration of 
other measures to mitigate the consequences of an accident occurring 
while a[n] ST is inoperable. Furthermore, administrative controls 
will be applied when exercising the 30-day CT extension and are 
adequate to maintain defense-in-depth and sufficient safety margins.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 26179]]

    Response: No.
    The proposed changes do not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. There [are] no design changes associated with the 
proposed changes. The changes to the CT do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter any of 
the assumptions made in the safety analysis. The changes to the CT 
do not affect the accident analysis directly; the CT is strictly 
tied to the PRA [probabilistic risk assessment] and the risk 
associated with the occurrence of a low-probability event during the 
limited time the component is unavailable.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. Neither the safety analyses nor the safety 
analysis acceptance criteria are impacted by these changes. The 
proposed changes will not result in plant operation in a 
configuration outside the current design basis. The proposed 
activities only involve changes to certain TS CTs.
    The proposed change does not involve a change to the plant 
design or operation and thus does not affect the design of the STs, 
the operation characteristics of the STs, the interfaces between the 
STs and other plant systems, or the function or reliability of the 
STs. Because the STs' performance and reliability will continue to 
be ensured by the proposed TS change, the proposed changes do not 
result in a reduction in the margin of safety.
    Therefore 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: David Terao.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: December 19, 2006.
    Brief description of amendments: The amendments requested would 
revise Technical Specification (TS) requirement 5.5.16, ``Containment 
Leakage Rate Testing Program,'' for consistency with the requirements 
of paragraph 50.55a(g)(4) of Title 10 of the Code of Federal 
Regulations (10 CFR) for components classified as Code Class CC.
    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[es] the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR [Part] 50, 
paragraph 55a(g)(4) for components classified as Code Class CC.
    The proposed change affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Containment Leakage Rate Testing Program. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The frequency of visual examinations of the concrete 
surfaces of the containment and the mode of operation during which 
those examinations are performed has no relationship to or adverse 
impact on the probability of any of the initiating events assumed in 
the accident analyses. The proposed change would allow visual 
examinations that are performed pursuant to NRC approved [American 
Society of Mechanical Engineers] (ASME) Section XI Code requirements 
(except where relief has been granted by the NRC) to meet the intent 
of visual examinations required by Regulatory Guide 1.163, without 
requiring additional visual examinations pursuant to the Regulatory 
Guide. The intent of early detection of deterioration will continue 
to be met by the more rigorous requirements of the Code required 
visual examinations. As such, the safety function of the containment 
as a fission product barrier is maintained.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. It does not involve the addition or removal of any 
equipment, or any design changes to the facility.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do[es] the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR [Part] 50, 
paragraph 55a(g)(4) for components classified as Code Class CC.
    The change affects the frequency of visual examinations that 
will be performed for the concrete surfaces containments. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Do[es] the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed change revises the Improved Standard Technical 
Specification Administrative Controls program requirements for 
consistency with the requirements of 10 CFR [Part] 50, paragraph 
55a(g)(4) for components classified as Code Class CC.
    The change affects the frequency of visual examinations that 
will be performed for the concrete surfaces containments. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The safety function of the containment as a fission product 
barrier will be maintained.
    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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Thomas Hiltz.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in

[[Page 26180]]

10 CFR Chapter I, which are set forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, 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, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of application for amendment: March 28, 2005, as supplemented 
by letters dated November 2, 2005, January 24, February 2, March 16, 
March 23, and March 28, 2007.
    Brief description of amendment: The amendment revises the Oyster 
Creek Licensing Basis in the area of radiological dose analyses for 
design-basis accidents using the alternative source terms depicted in 
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for 
Evaluating Design Basis Accidents at Nuclear Power Reactors.'' 
Additionally, the amendment revises the Oyster Creek Technical 
Specifications (TSs) consistent with the amended design-basis.
    Date of Issuance: April 26, 2007.
    Effective date: As of the date of Issuance to be implemented within 
60 days.
    Amendment No.: 262.
    Facility Operating License No. DPR-16: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24646). The supplemental letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the Nuclear Regulatory 
Commission (NRC) staff's original proposed to significant hazards 
consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 26, 2007.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

    Date of application for amendment: June 1, 2006, as supplemented by 
letters dated November 20, 2006, and February 22, 2007.
    Brief description of amendment: The amendment revises Surveillance 
Requirement 3.5.2 in the HBRSEP2 Technical Specifications.
    Date of issuance: April 4, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 213.
    Renewed Facility Operating License No. DPR-23: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: December 19, 2006 (71 
FR 75992). The supplemental letters provided additional information 
that was within the scope of the original notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 4, 2007.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: November 27, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.9 to relocate the specific American Society of 
Testing and Materials (ASTM) Standard from the Administrative Controls 
Section of TS to a licensee-controlled document. Also, the revision to 
TS 5.5.9 allows the performance of an alternate water and sediment 
content test to establish the acceptability of new fuel oil prior to 
addition to the storage tank has been added to the clear and bright 
test.
    Date of issuance: April 12, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 174.
    Facility Operating License No. NPF-43: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: January 3, 2007 (72 FR 
149).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 13, 2006.
    Brief description of amendment: The amendment revised Facility 
Operating License No. NPF-58 by deleting License Condition 2.F, which 
specifies reporting of 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).
    Date of issuance: April 19, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 140.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: November 21, 2006 (71 
FR 67394).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2007.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of application for amendment: August 31, 2006, as supplemented 
on December 15, 2006, and March 1 and April 4, 2007.
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of Renewed Facility Operating License No. DPR-
20 to Entergy Nuclear Palisades, LLC, as

[[Page 26181]]

owner, and Entergy Nuclear Operations, Inc., as operator, as approved 
by Order of the Commission dated April 6, 2007, and as revised on April 
10, 2007.
    Date of issuance: April 11, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 224.
    Facility Operating License No. DPR-20: Amendment revised the 
Renewed Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 16, 2006 (71 
FR 66805).
    The December 15, 2006, and March 1 and April 4, 2007, supplemental 
letters contained clarifying information and did not expand the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 2007, as revised on April 10, 
2007.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of application for amendment: December 20, 2006.
    Brief description of amendment: The amendment revises License 
Condition 2.B.3(c) to allow the receipt, possession, and use of 
byproduct, source, or special nuclear material without restriction to 
amount or atomic number, for sample analysis or instrument calibration 
or associated with radioactive apparatus or components.
    Date of issuance: April 17, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 39.
    Facility