[Federal Register Volume 70, Number 108 (Tuesday, June 7, 2005)]
[Notices]
[Pages 33210-33225]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-2848]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

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 May 13, 2005 to May 25, 2005. The last 
biweekly notice was published on May 24, 2005 (70 FR 29785).

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

[[Page 33211]]

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, Rules and 
Directives 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    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, [email protected]; 
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 
[email protected]. 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, http://

[[Page 33212]]

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 [email protected].

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

    Date of amendment request: May 18, 2005.
    Description of amendment request: The proposed amendment would 
revise Fermi 2 Technical Specifications (TSs) to add Actions to 
Limiting Condition for Operation (LCO) 3.8.1, ``AC Sources--
Operating,'' for one offsite circuit inoperable, for two offsite 
circuits inoperable, and for one offsite circuit and one or both 
emergency diesel generators (EDGs) in one Division inoperable, in 
accordance with Regulatory Guide 1.93, ``Availability of Electric Power 
Sources.'' The current Fermi 2 TSs contain only a single Action for one 
or two offsite circuits inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to replace the existing LCO 3.8.1 Action C 
for one or two offsite circuits inoperable with a required 
Completion Time of 12 hours to be in MODE 3, and 36 hours to be in 
MODE 4, with new Actions C, D, and E to allow a single offsite 
circuit to be inoperable for up to 72 hours, two offsite circuits to 
be inoperable for up to 24 hours, and one offsite circuit and one or 
both EDGs in one Division to be inoperable for up to 12 hours, 
provided other Required Actions are taken is consistent with the 
NUREG 1433, ``Standard Technical Specifications General Electric 
Plants, BWR/4,'' criteria, and with the guidelines in Regulatory 
Guide 1.93. There is no change in plant design, and [Title 10 of the 
Code of Federal Regulations (10 CFR)] 10 CFR 50, Appendix A, General 
Design Criteria 17, ``Electric Power Systems'' will continue to be 
met. Increasing the Completion Times for inoperable offsite circuits 
will not significantly increase the potential for a loss of offsite 
power. This is due to the redundancy and diversity of the offsite 
electrical configuration at Fermi 2. Inoperability of an offsite 
circuit does slightly increase the potential for a loss of 
divisional power. The probability of losing the opposite division of 
offsite power in this condition is extremely small due to the 
physical separation of the offsite power sources that feed Fermi 2. 
Furthermore, the 10 CFR 50.65(a)(4) program monitors the condition 
of the offsite electrical system and switchyard configuration for 
each entry into the extended completion time to ensure that there is 
no significant increase in the probability or consequences of an 
accident.
    The proposed change does not alter the operation of any plant 
equipment assumed to function in response to an analyzed event or 
otherwise increase its failure probability. Therefore, this change 
does not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not alter the design, configuration, or 
method of operation of the plant. It simply provides longer 
Completion Times for inoperable offsite circuits. No physical or 
operational changes to the components of the A. C. power systems are 
being made by this change; therefore, no new system interactions are 
being created. The proposed change does not produce any parameters 
or conditions that could contribute to the initiation of accidents 
different from those already evaluated. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change will replace the existing LCO 3.8.1 Action C 
for one or two offsite circuits inoperable with a required 
Completion Time of 12 hours to be in MODE 3, and 36 hours to be in 
MODE 4, with new Actions C, D, and E to allow a single offsite 
circuit to be inoperable for up to 72 hours, two offsite circuits to 
be inoperable for up to 24 hours, and one offsite circuit and one or 
both EDGs in one Division to be inoperable for up to 12 hours, 
provided other Required Actions are taken. This change is consistent 
with NUREG 1433, ``Standard Technical Specifications General 
Electric Plants, BWR/4,'' and with the guidelines in Regulatory 
Guide 1.93. The proposed change does not affect any analysis that is 
used to establish safety margins, nor does it alter the design, 
configuration, or method of operation of the plant. 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: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Section Chief: L. Raghavan.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: April 19, 2005.
    Description of amendment request: The proposed amendment would 
revise technical specifications (TS) testing frequency for the 
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' 
Specifically, the proposed change would revise the frequency for SR 
3.1.4.2, Control Rod Scram Time Testing, from ``120 days cumulative 
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the 
model NSHC determination in its application dated April 19, 2005.
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in 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 extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. 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 extends the frequency for testing control 
rod scram time

[[Page 33213]]

testing from every 120 days of cumulative Mode 1 operation to 200 
days of cumulative Mode 1 operation. The proposed change continues 
to test the control rod scram time to ensure the assumptions in the 
safety analysis are protected. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

    Based on the above, the proposed change presents no significant 
hazards consideration under the standards set forth in 10 CFR 50.92(c), 
and accordingly, a finding of ``no significant hazards consideration'' 
is justified.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point 
Nuclear Generating Unit Nos. 2 and 3 (IP2 and 3), Westchester County, 
New York

    Date of amendment request: April 22, 2005.
    Description of amendment request: The amendments would revise the 
surveillance requirements (SRs) for Technical Specification (TS) 3.3.5, 
``Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.'' 
Specifically, a note would be added to IP2 TS SR 3.3.5.2 to indicate 
that the verification of the setpoint is not required for the 480 volt 
(V) bus degraded voltage function when performing the trip actuating 
device operational test (TADOT). A similar note would be added to IP3 
TS SR 3.3.5.1 for the 480V degraded voltage and undervoltage functions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously 
evaluated[?]
    Response: No.
    The proposed change adds a note to indicate that the IP2 and IP3 
degraded voltage relays and the IP3 undervoltage relays do not 
require setpoint verification when the TADOT required by TS 
surveillances is performed on a monthly basis. Setpoint verification 
of these relays occurs as part of the channel calibration that is 
performed at either an 18 month or a 24 month frequency. These 
relays are used to sense either degraded voltage or undervoltage on 
the 480 volt safety related buses and to initiate the start of the 
EDG [emergency diesel generator] for all events where the loss of 
offsite power is postulated. This function has no effect on the 
probability of an accident previously evaluated since it is not 
associated with the initiation of any accident. The relay setpoint 
verification frequency of 18 or 24 months has no significant effect 
on the consequences of an accident because the relays are intended 
to be calibrated on this frequency. This frequency of calibration is 
based on operating experience, and is consistent with industry 
practice. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change adds a note to indicate that the IP2 and IP3 
degraded voltage relays and the IP3 undervoltage relays do not 
require setpoint verification when the TADOT required by TS 
surveillances is performed on a monthly basis. This effectively 
changes the frequency required by the surveillance requirement from 
31 days to either 18 months or 24 months. The change does not affect 
the function of the relays or otherwise affect the design and 
operation of plant systems and components and therefore no new 
accident scenarios would be created. The change does not affect the 
manner is which equipment is operated but does affect the manner in 
which it is maintained by extending the frequency for setpoint 
verification. The frequency change continues to provide adequate 
verification of the operability of equipment and limits the time 
which the relay function is inoperable or degraded while performing 
verification. Therefore, no new failure modes are being introduced 
that could lead to different accidents.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds a note to indicate that the IP2 and IP3 
degraded voltage relays and the IP3 undervoltage relays do not 
require setpoint verification when the TADOT required by TS 
surveillances is performed on a monthly basis. Setpoint verification 
of these relays occurs as part of the channel calibration that is 
performed at either an 18 month or a 24 month frequency. The margin 
associated with these relays is the assurance that these relays will 
properly sense either degraded voltage or undervoltage on the 480 
volt safety related buses and to initiate the start of the EDG for 
all events where the loss of offsite power is postulated. The 
proposed frequency of calibration is based on operating experience, 
and is consistent with industry practice. These indicate that 
setpoint verification at 18 month or 24 month [frequency] is 
adequate to assure performance of the function. Verification of 
setpoints on a monthly basis either degrades the reliability of the 
function or makes it inoperable. 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. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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: April 13, 2005.
    Description of amendment request: The proposed amendments would 
extend the completion time (CT) for required Action A.1, ``Restore 
Residual Heat Removal Service Water (RHRSW) subsystem to OPERABLE 
status,'' associated with Technical Specification (TS) Section 3.7.1 
from 7 days to 10 days. This proposed change would only be used during 
the upcoming Unit 1 2006 refueling outage. The establishment of a 6 day 
(for Division 2 core standby cooling system (CSCS) maintenance) or 10 
day (for Division 1 CSCS maintenance ) CT for TS Section 3.7.2 when one 
or more required diesel generator cooling water (DGCW) subsystem(s) are 
inoperable. This proposed change will only be used during each of the 
upcoming Unit 1 2006, and Unit 2 2007, refueling outages, and during 
the subsequent Unit 1 2008, refueling outage. An extension of the CT 
for required Action C.4, ``Restore required Diesel Generator (DG) to 
OPERABLE status,'' associated with TS Section 3.8.1 from 72 hours to 6 
days. This proposed change will only be used during the upcoming Unit 2 
2007 refueling outage, and during subsequent Unit 1, 2008, refueling 
outage. An extension of the CT for required Action F.1, ``Restore one 
required Diesel Generator (DG) to OPERABLE status,'' associated with TS 
Section 3.8.1 from 2 hours to 6 days. This proposed change will only be 
used during the upcoming Unit 2, 2007, refueling outage, and during 
subsequent Unit 1, 2008, refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 33214]]

    The proposed changes have been evaluated using the risk-informed 
processes described in RG [Regulatory Guide] 1.174, ``An Approach 
for Using Probabilistic Risk Assessment in Risk-Informed Decisions 
on Plant-Specific Changes to the Licensing Basis,'' dated July 1998, 
and RG 1.177, ``An Approach for Plant-Specific, Risk-Informed 
Decision Making: Technical Specifications,'' dated August 1998. The 
risk associated with the proposed change was found to be acceptable.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
does not have a detrimental impact on the integrity of any plant 
structure, system, or component that initiates an analyzed event. No 
active or passive failure mechanisms that could lead to an accident 
are affected. Non-code line stops required to isolate the Unit 1 
portion of the common discharge header from the Unit 2 portion of 
the header during the specified CSCS maintenance will maintain the 
availability of the online unit's Division 2 CSCS system. The non-
code line stops being used to isolate the system during the 
specified refueling outages are being designed to the same pressure 
rating and seismic requirements as the CSCS piping.
    Redundancy is provided by designing the CSCS system as multiple 
independent subsystems. Separation between subsystems assures that 
no single failure can affect more than one subsystem. Therefore, 
assuming a single failure in any subsystem including the subsystem 
shared between units, two subsystems in each unit will remain 
unaffected. These two subsystems can supply the minimum required 
cooling water for safe shutdown of a unit or mitigate the 
consequences of an accident.
    The proposed limited use of increased CT's of the operating 
unit's CSCS system maintains the design basis assumptions; 
therefore, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change involves the temporary installation of new 
equipment (mechanical line stops) that will be designed and 
installed to the same pressure rating and seismic design as the CSCS 
piping. The currently installed equipment will not be operated in a 
new or different manner. No new or different system interactions are 
created and no new processes are introduced. The proposed changes 
will 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. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not alter any existing setpoints at 
which protective actions are initiated and no new setpoints or 
protective actions are introduced. The design and operation of the 
CSCS system remains unchanged. The risk assessment with the proposed 
increase in the CTs for TS 3.7.1, TS 3.7.2, and TS 3.8.1 were 
evaluated using the risk-informed processes described in RG 1.174, 
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing 
Basis,'' dated July 1998, and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decision Making: Technical Specifications,'' 
dated August 1998. The risk was shown to be acceptable. Based on 
this evaluation, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief : Gene Y. Suh.

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

    Date of amendment request: April 11, 2005.
    Description of amendment request: The proposed amendment would 
revise the BVPS-2 Technical Specification (TS) 3.4.5 to change the 
scope of the steam generator (SG) tubesheet examinations required in 
the SG tubesheet region by using the F* inspection methodology. 
Specifically, the proposed amendment would alter the tube inspection to 
exclude the portion of the SG tube within the tubesheet below the F* 
distance and to exclude the tube-to-tubesheet weld, by crediting the 
methodology described in Westinghouse Topical Report, WCAP-16385, 
Revision 1. The F* distance is the distance from the top of the 
tubesheet to the bottom of the F* length (the maximum length of tubing 
below the bottom of the roll transition (BRT) which must be 
demonstrated to be non-degraded and which is defined as 1.97 inches on 
the hot leg side) plus the distance to the BRT and non-destructive 
examination uncertainties. The licensee's proposed amendment also would 
revise the TS requirements to require tubes with service-induced 
degradation identified in the F* distance or less than or equal to 3.0 
inches below the top of the tubesheet, whichever is greater, to be 
repaired or removed from service upon detection. The TS Index, affected 
TS pages and Bases would also be revised and repaginated as necessary 
to reflect the proposed TS change.
    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?
    No. The proposed change modifies the BVPS Unit 2 TSs to 
incorporate steam generator tube inspection scope based on WCAP-
16385, Revision 1. Of the various accidents previously evaluated in 
the BVPS Unit 2 Updated Final Safety Analysis Report (UFSAR), the 
proposed changes only affect the steam generator tube rupture (SGTR) 
event evaluation and the postulated steam line break (SLB) accident 
evaluation. Loss-of-coolant accident (LOCA) conditions cause a 
compressive axial load to act on the tube. Therefore, since the LOCA 
tends to force the tube into the tubesheet rather than pull it out, 
it is not a factor in this amendment request. Another faulted load 
consideration is a safe shutdown earthquake (SSE); however, the 
seismic analysis of Model 51M SGs has shown that axial loading of 
the tubes is negligible during an SSE.
    For the SGTR event, the required structural margins of the steam 
generator tubes will be maintained by the presence of the tubesheet. 
Tube rupture is precluded for cracks in the tube expansion region 
due to the constraint provided by the tubesheet. Therefore, 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR 
[pressurized-water reactor] Steam Generator Tubes,'' margins against 
burst are maintained for both normal and postulated accident 
conditions.
    The F* length supplies the necessary resistive force to preclude 
pullout loads under both normal operating and accident conditions. 
The contact pressure results from the tube expansion process used 
during manufacturing and from the differential pressure between the 
primary and secondary side. The proposed changes do not affect other 
systems, structures, components or operational features. Therefore, 
the proposed change results in no significant increase in the 
probability of the occurrence of an SGTR or SLB accident.
    The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary 
leakage flow through a postulated broken tube is not affected by the 
proposed change since the tubesheet enhances the tube integrity in 
the region of the expansion by precluding tube deformation beyond 
its initial expanded outside diameter. The resistance to both tube 
rupture and collapse is strengthened by the tubesheet in that 
region. At normal operating pressures, leakage from primary water 
stress corrosion cracking (PWSCC) below the F* length is limited by 
both the tube-to-tubesheet crevice and the limited crack opening 
permitted by

[[Page 33215]]

the tubesheet constraint. Consequently, negligible normal operating 
leakage is expected from cracks within the tubesheet region.
    SLB leakage is limited by leakage flow restrictions resulting 
from the crack and tube-to-tubesheet contact pressures that provide 
a restricted leakage path above the indications and also limit the 
degree of crack face opening compared to free span indications. The 
total leakage (i.e., the combined leakage for all such tubes) meets 
the industry performance criterion, plus the combined leakage 
developed by any other alternate repair criteria, and will be 
maintained below the maximum allowable SLB leak rate limit, such 
that off-site doses are maintained less than 10 CFR [Part] 100 
guideline values and the limits evaluated in the BVPS Unit 2 UFSAR.
    Therefore, based on the above evaluation, the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed changes do not introduce any changes or 
mechanisms that create the possibility of a new or different kind of 
accident. Tube bundle integrity will continue to be maintained for 
all plant conditions upon implementation of the F* methodology.
    The proposed changes do not introduce any new equipment or any 
change to existing equipment. No new effects on existing equipment 
are created nor are any new malfunctions introduced.
    Therefore, 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?
    No. The proposed changes maintain the required structural 
margins of the steam generator tubes for both normal and accident 
conditions, including the planned uprated power level of 2910 Mwt. 
NRC [Nuclear Regulatory Commission] Regulatory Guide (RG) 1.121 is 
used as the basis in the development of the F* methodology for 
determining that steam generator tube integrity considerations are 
maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC staff for meeting General Design Criteria 14, 
15, 31, and 32 by reducing the probability and consequences of an 
SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions of tube wall degradation beyond which tubes with 
unacceptable cracking, as established by inservice inspection, 
should be removed from service or repaired, the probability and 
consequences of an SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For primarily axially oriented cracking located within the 
tubesheet, tube burst is precluded due to the presence of the 
tubesheet. WCAP-16385, Revision 1, defines a length, F*, of 
degradation-free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure-induced forces (with 
applicable safety factors applied). Application of the F* criteria 
will preclude unacceptable primary-to-secondary leakage during all 
plant conditions. The methodology for determining leakage provides 
for large margins between calculated and actual leakage values in 
the F* criteria.
    Plugging of the steam generator tubes reduces the reactor 
coolant flow margin for core cooling. Implementation of F* 
methodology at Beaver Valley Unit 2 will result in maintaining the 
margin of flow that may have otherwise been reduced by tube 
plugging.
    Based on the above, it is concluded that the proposed changes do 
not result in a significant reduction of margin with respect to 
plant safety as defined in the Final Safety Analysis Report Update 
or bases of the plant Technical Specifications.

    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit No. 2 (SL2), St. Lucie County, Florida

    Date of amendment request: March 31, 2005.
    Description of amendment request: The proposed amendment would 
revise Administrative Technical Specification Section 6.8.4.h, 
``Containment Leakage Rate Testing Program,'' to allow a one-time 
extension of the currently approved 15-year test interval to 
approximately 15.5 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed amendment of the Technical Specifications adds a one-
time extension to the current surveillance interval for Type A 
testing (ILRT [integrated leak rate testing]). The current test 
interval of 15 years from the last Type A test would be extended to 
end prior to startup from the SL2-17 refueling. This is anticipated 
to be an approximately six-month addition to the 15 year interval. 
The proposed extension to the Type A testing interval does not 
significantly increase the probability of an accident previously 
evaluated since the containment Type A test is not a modification, 
nor a change in the way that plant systems, structures or components 
(SSC) are operated, and is not an activity that could lead to 
equipment failure or accident initiation. The proposed extension of 
the test interval does not involve a significant increase in the 
consequences of an accident since research documented in NUREG-1493 
has found that generically, very few potential leak paths are not 
identified with Type B and C tests (LLRT [local leak-rate test]). 
The Type B and C testing are unaffected by this proposed change. The 
NUREG concluded that an increase in the Type A test interval to 
twenty years resulted in an imperceptible increase in risk. St. 
Lucie Unit 2 provides a high degree of assurance through testing and 
inspection that the containment will not degrade in a manner only 
detectable by Type A testing. Inspections required by the ASME 
[American Society of Mechanical Engineers] Code, the containment 
leakage rate testing program, the plant protective coatings program, 
and Maintenance Rule are performed in order to identify indications 
of containment degradation that could affect leak tightness. Type B 
and C testing required by 10 CFR 50, Appendix J, are not affected by 
this proposed extension to the Type A test interval and will 
identify openings in containment penetrations that would otherwise 
require a Type A test.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed change does not result in facility operation that 
would create the possibility of a new or different kind of accident 
from any accident previously evaluated. The proposed extension to 
Type A testing does not create a new or different type of accident 
for St. Lucie because no physical plant changes are made and no 
compensatory measures are being imposed that could potentially lead 
to a failure. There are no operational changes that could introduce 
a new failure mode or create a new or different kind of accident. 
The proposed change only adds an extension to the current interval 
for Type A testing and does not change implementation aspects of the 
test.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed change would not result in operation of the 
facility involving a significant reduction in a margin of safety. 
The proposed license amendment adds a one-time extension to the 
current interval for Type A testing (ILRT). The current one-time 
test interval of 15 years from the last Type A test would be 
extended to end prior to startup from the SL2-17 refueling outage. 
This is anticipated to be an approximately six month addition to the 
15 year interval.

[[Page 33216]]

    The NUREG-1493 generic study of the effects of extending the 
Type A test interval out to 20 years concluded that there is an 
imperceptible increase in plant risk. A plant specific risk 
calculation obtained results consistent with the generic conclusions 
regarding risk which show a slight but negligible increase in risk. 
Inspections required by the ASME code and maintenance rule are 
performed to ensure that the containment will not degrade in a 
manner that is only detectable by Type A testing (ILRT).

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: April 13, 2005.
    Description of amendment request: The proposed amendment would 
incorporate several Technical Specification Task Force (TSTF) changes 
to the licensee's Technical Specifications (TSs). The specific TSTF 
changes that would be incorporated are:
    1. TSTF-222-A, Revision 1, ``Control Rod Scram Time Testing''--This 
change modifies TS Section 3.1.4, ``Control Rod Scram Times,'' to 
clarify that control rod scram time testing is required only for core 
cells in which work on the control rod or drive has been performed or 
fuel has been moved or replaced.
    2. TSTF-275-A, Revision 0, ``Clarify Requirement for EDG [emergency 
diesel generator] start signal on RPV [reactor pressure vessel] Level--
Low, Low, Low during RPV cavity flood-up''--This change modifies the TS 
Section 3.3.5.1, ``ECCS [emergency core cooling system] 
Instrumentation,'' to clarify that the ECCS initiation instrumentation, 
identified as being required in modes 4 and 5, is required to be 
operable only when the associated ECCS subsystems are required to be 
operable as defined in limiting condition of operation (LCO) 3.5.2, 
``ECCS--Shutdown.''
    3. TSTF-300-A, Revision 0, ``Eliminate DG [diesel generator] LOCA 
[loss-of-coolant accident]--Start SRs [surveillance requirements] while 
in S/D [shutdown] when no ECCS is Required''--This change modifies the 
TS Section 3.8.2, ``AC [alternating current] Sources--Shutdown,'' to 
add an additional note to the surveillance that verifies automatic 
start of the emergency diesel generators and automatic load shedding 
from the emergency buses, is considered to be met without the ECCS 
initiation signals operable when ECCS initiation signals are not 
required to be operable per Table 3.3.5.1-1, ECCS Instrumentation.
    4. TSTF-225, Revision 2, ``Fuel movement with inoperable refueling 
equipment interlocks''--This change modifies TS Section 3.9.1, 
``Refueling Equipment Interlocks,'' to add required actions to allow 
insertion of a control rod withdrawal block and verification that all 
control rods are fully inserted as alternate actions to suspending in-
vessel fuel movement in the event that one or more required refueling 
equipment interlocks are inoperable.
    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.
    1. Revision of CNS [Cooper Nuclear Station] TS SR 3.1.4.1 and SR 
3.1.4.4. The frequency at which control rod scram time is verified 
is not a precursor of an accident. A scram time slower than required 
might result in an increase in the consequences of an accident. 
However, revising the frequency for verifying the scram time of the 
control rods does not impact the scram time. Verifying that the 
scram time is acceptable will continue to be required prior to plant 
startup following fuel movement or work on the control rods or 
control rod drive system. Therefore, revising the frequency for 
verifying insertion time to clarify when it is required does not 
involve a significant increase in the probability of an accident or 
an increase in the consequences of an accident.
    2. Revision of TS Table 3.3.5.1-1. Clarifying when certain ECCS 
instrumentation must be operable with the plant shut down will not 
increase either the probability of an accident or the consequences 
of the accident. The ECCS instrumentation is required to be operable 
only when the associated ECCS subsystems are required to be 
operable. This continues to ensure that the instrumentation will be 
operable when it is required.
    3. Revision of TS SR 3.8.2.1. The frequency of verifying certain 
actions by surveillances is not a precursor to accidents. Clarifying 
that the actions required in response to an ECCS initiation signal 
are not required when the ECCS initiation signals are not required 
to be operable does not result in increased probability of an 
accident or increased consequences of an accident. Not requiring 
that a DG automatically start in response to the ECCS initiation 
signal when the ECCS subsystems that are supported by the DG are not 
required to be operable does not reduce the required ECCS 
protection.
    4. Revision of TS 3.9.1., Condition A Required Action. The 
actions taken when a refueling equipment interlock is inoperable are 
not initiators of any accident previously evaluated. The level of 
protection against withdrawing a control rod during the insertion of 
a fuel assembly or loading a fuel assembly into the vessel with a 
control rod withdrawn, provided by the proposed alternate Required 
Actions, is equivalent to that provided by the current Required 
Action. The radiological consequences of an accident described in 
the Updated Safety Analysis Report (USAR) while taking the proposed 
alternate Required Actions are not different from the consequences 
of an accident under the current Required Actions.
    Based on the above NPPD [Nebraska Public Power District] 
concludes that 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?
    Response: No.
    The proposed changes to the CNS operating license involve 
revisions to the requirements for when certain surveillances are to 
be performed (change no. 1 and no. 3), clarification of when ECCS 
instrumentation is required to be operable (change no. 2), and 
addition of alternative Required Actions if certain plant components 
are inoperable (change no. 4). These changes will not result in 
revision of plant design, physical alteration of a plant structure, 
system, or component (SSC), or installation of a new or different 
type of equipment. The changes do not involve any revision of how 
the plant, an SSC, or a refueling equipment interlock, are operated. 
Based on this, the proposed changes do not create the possibility of 
a new or different kind of accident.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    1. Revision of CNS TS SR 3.1.4.1 and SR 3.1.4.4. Sufficiently 
rapid insertion of control rods following certain accidents (scram 
time) will prevent fuel damage, and thereby maintain a margin of 
safety to fuel damage. No change is being made to the required 
insertion rate specified in plant technical specifications. 
Clarifying when control rod insertion times must be verified 
following movement of fuel assemblies, without actually changing the 
requirement (verification of insertion times will continue to be 
required whenever work that might impact the rod insertion time is 
done), does not reduce the margin of safety related to fuel damage.
    2. Revision of TS Table 3.3.5.1-1. Clarifying when certain ECCS 
instrumentation is required to be operable when CNS is in a shutdown 
mode does not change the requirement. Not requiring ECCS signals 
that initiate a DG to be operable when the ECCS subsystems that are 
supported by

[[Page 33217]]

the DG are not required to be operable does not result in a 
reduction of a margin of safety for the safety related equipment 
that is required to be operable.
    3. Revision of TS SR 3.8.2.1. Clarifying that automatic start of 
the DGs in response to the ECCS initiation signal is not required 
when the ECCS subsystems that are supported by the DG are not 
required to be operable does not result in a reduction in a margin 
of safety.
    4. Revision of TS 3.9.1, Condition A Required Action. The 
proposed alternate Required Actions to be taken when a refueling 
interlock is inoperable provide a level of protection against 
inadvertent criticality while inserting or moving fuel in the 
reactor vessel that is equivalent to the level provided by the 
current Required Action. As a result, the proposed alternate 
Required Actions do not result in a significant reduction in a 
margin of safety related to protection against inadvertent 
criticality when inserting or moving fuel assemblies.
    Based on the above NPPD concludes that the proposed changes do 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: David Terao.

PSEG Nuclear, LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.1.3.1, ``Control Rod 
Operability,'' such that scram discharge volume (SDV) vent or drain 
lines with inoperable valves would be isolated instead of requiring 
that the valve be restored to Operable status or the unit be placed in 
Hot Shutdown within 12 hours.
    The NRC staff issued a Notice of Opportunity for Comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a Notice of Availability of the models for referencing license 
amendment applications in the Federal Register on April 15, 2003 (68 FR 
18294). The licensee affirmed the applicability of the model NSHC 
determination (modified slightly to address plant-specific TS format) 
in its application dated February 25, 2005.
    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.
    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with inoperable valves instead or requiring the valves to be 
restored to operable status or the unit be in hot shutdown within 12 
hours. With SDV vent or drain valves inoperable in one or more 
lines, the isolation function would be maintained since the 
redundant valve in the affected line would perform its safety 
function of isolating the SDV. Following the completion of the 
required action, the isolation function is fulfilled since the 
associated line is isolated. The ability to vent and drain the SDV 
is maintained and controlled through administrative controls. This 
requirement assures the reactor protection system is not adversely 
affected by the inoperable valves. With the safety functions of the 
valves being maintained, the probability or consequences of an 
accident previously evaluated are not significantly increased.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in [a] margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDV is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of the SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    Based on the reasoning presented above, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: March 10, 2005.
    Description of amendment request: The amendment would revise 
Technical Specification Section 5.5.15, ``Containment Leakage Rate 
Testing Program,'' to allow a one-time extension of the interval 
between the Type A, integrated leakage rate tests (ILRTs), from 10 
years to no more than 15 years.
    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 to Technical Specification 5.5.15, 
Containment Leakage Rate Testing Program, involves a one-time 
extension to the current interval for Type A containment testing. 
The current test interval of ten (10) years would be extended on a 
one-time basis to no longer than fifteen (15) years from the last 
Type A test.
    The proposed Technical Specification change does not involve a 
physical change to the plant or a change in the manner which the 
plant is operated or controlled. The reactor containment is designed 
to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. As such the reactor containment itself and the 
testing requirements invoked to periodically demonstrate the 
integrity of the reactor containment exist to ensure the plant's 
ability to mitigate the consequences of an accident, and do not 
involve the prevention or identification of any precursors of an 
accident.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG-1493, that Type B and C 
containment leakage tests have identified a very large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is very 
small. The Ginna ILRT test history supports this conclusion. In 
NUREG-1493 Section 10, Summary of Technical Findings, it is 
concluded, in part, that reducing the frequency of Type A 
containment leak tests to once per twenty (20) years leads to an 
imperceptible increase in risk.

[[Page 33218]]

    The proposed change does not result in an increase in core 
damage frequency since the containment system is used for mitigation 
purposes only. Containment Leakage Rate Testing Program local leak 
rate test requirements and administrative controls such as design 
change control, ASME [American Society of Mechanical Engineers] 
Section XI Inservice Inspection (ISI) Program Containment Repair and 
Replacement Program and procedural requirements for system 
restoration ensure that containment integrity is not degraded by 
plant modifications or maintenance activities. The design and 
construction requirements of the reactor containment itself combined 
with the containment inspections performed in accordance with the 
ASME Section XI Inservice Inspection (ISI) Program Containment 
Program, Boric Acid Corrosion Program, inspections in accordance 
with Regulatory Guide 1.163 position C.3 and the Maintenance Rule 
serve to provide a high degree of assurance that the containment 
will not degrade in a manner that is detectable only by Type A 
testing.
    Therefore, the proposed Technical Specification change does not 
involve a significant increase in the 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 to Technical Specification 5.5.15 involves a 
one-time extension to the current interval for Type A containment 
testing. The reactor containment and the testing requirements 
invoked to periodically demonstrate the integrity of the reactor 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident and do not involve the prevention or 
identification of any precursors of an accident. The proposed 
Technical Specification change does not involve a physical change to 
the plant (i.e., no new or different type of equipment will be 
installed) or changes in the methods in which the plant is operated 
or controlled.
    Therefore, the proposed Technical Specification change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    The proposed change to Technical Specifications involves a one-
time extension to the current interval for Type A containment 
testing. The proposed Technical Specification change does not alter 
the manner in which safety limits, limiting safety system set 
points, or limiting conditions for operation are determined. The 
specific requirements and conditions of the Primary Containment 
Leakage Rate Testing Program, as defined in Technical 
Specifications, exist to ensure that the degree of reactor 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leakage rate limit specified by Technical Specifications 
is maintained. The proposed change involves only the extension of 
the interval between Type A containment leakage tests. Type B and C 
containment leakage tests will continue to be performed at the 
frequency currently required by plant Technical Specifications.
    Ginna and industry experience strongly supports the conclusion 
that Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with the ASME Section XI 
Inservice Inspection (ISI) Program Containment Program, Boric Acid 
Corrosion Program, inspections in accordance with Regulatory Guide 
1.163 position C.3 and the Maintenance Rule serve to provide a high 
degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing. The combination of 
these factors ensures that the margin of safety that is inherent in 
plant safety analysis is maintained.
    Therefore, the proposed Technical Specification change 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: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: April 29, 2005.
    Description of amendment request: The amendment would revise 
Technical Specification Section 3.7.3, ``Main Feedwater Regulating 
Valves (MFRVs), Associated Bypass Valves, and Main Feedwater Pump 
Discharge Valves (MFPDVs),'' to allow the use of the main feedwater 
isolation valves in lieu of the main feedwater pump discharge valves to 
provide isolation capability to the steam generators in the event of a 
steam line break.
    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 proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes involve a modification to the plant 
configuration to ensure the acceptability of containment response 
for Steam Line Breaks (SLB) inside containment.
    The changes have also been evaluated to ensure the core response 
for steam system piping breaks remains acceptable. The changes to 
the Technical Specifications (TS) are necessary to properly 
accommodate the changes in plant configuration and ensure proper 
testing of the modified components.
    The proposed changes do not adversely affect accident initiators 
or precursors nor significantly alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes do not 
adversely 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. Further, the proposed changes do not increase 
the types and amounts of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposures. The proposed changes cannot 
affect the probability of an accident occurring since they reflect a 
change in plant design consistent with current design which is not 
an accident initiator. The proposed changes cannot increase the 
consequences of postulated accidents since they reflect a change in 
plant design that will continue to mitigate the effects of feedwater 
addition to a faulted steam generator for a main steam line break 
inside containment.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve a modification to the plant 
configuration to ensure the acceptability of containment response 
for Steam Line Breaks (SLB) inside containment. The changes have 
also been evaluated to ensure the core response for steam system 
piping breaks remains acceptable. The changes to the Technical 
Specifications (TS) are necessary to properly accommodate the 
changes in plant configuration and ensure proper testing of the 
modified components.
    The change in plant configuration significantly reduces the 
available water volume and therefore the mass and energy released to 
the containment in the event of an SLB with failure of a feedwater 
regulating valve. Existing feedwater flow paths or piping are not 
significantly altered. An existing manual valve in the flow path to 
each steam generator is utilized as the main feedwater isolation 
valve by the addition of an air actuator to provide automatic 
isolation capability. The changes do not involve a significant 
change in the methods governing normal plant operation. The TS 
changes modify the limiting condition for operation, required action 
statements, associated completion times and surveillance 
requirements to those that are consistent with those previously 
approved for Westinghouse

[[Page 33219]]

plants in the Standard Technical Specifications found in NUREG-1431. 
The proposed TS changes do not create the possibility of a new or 
different [kind] of accident from those previously evaluated since 
they reflect a design change that will accomplish the same feedwater 
isolation function as previously performed by the main feedwater 
pump discharge isolation valves with no significant change to the 
manner in which the feedwater system operates.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes involve a modification to the plant 
configuration to ensure the acceptability of containment response 
for Steam Line Breaks (SLB) inside containment. The changes have 
also been evaluated to ensure the core response for steam system 
piping breaks remains acceptable. The changes to the Technical 
Specifications (TS) are necessary to properly accommodate the 
changes in plant configuration and ensure proper testing of the 
modified components.
    The level of safety of facility operation is unaffected by the 
proposed changes since there is no change in the intent of the TS 
requirements of assuring proper main feedwater isolation in the 
event of a steam line break inside containment. The response of the 
plant systems to accidents and transients reported in the Updated 
Final Safety Analysis Report (UFSAR) is not adversely affected by 
this change. Therefore, the capability to satisfy accident analysis 
acceptance criteria is not adversely affected. The TS changes modify 
the limiting condition for operation, required action statements, 
associated completion times and surveillance requirements to those 
that are consistent with those previously approved for Westinghouse 
plants in the Standard Technical Specifications found in NUREG-1431. 
The proposed TS changes do not involve a significant reduction in 
[a] margin of safety since they are based upon a modification that 
will maintain [a] margin of safety with respect to feedwater 
addition for a main steam line break inside containment to the 
previously analyzed condition. Therefore, the changes do not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: April 29, 2005.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.5.1, ``Accumulators,'' and TS 3.5.4, 
``Refueling Water Storage Tank (RWST),'' to reflect the results of 
revised analyses performed to accommodate a planned power uprate for 
the facility and revise TS 5.6.5, ``Core Operating Limits Report 
(COLR),'' to permit the use of NRC-approved methodology for large-break 
and small-break loss-of-coolant accidents (LOCAs).
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes include revising accumulator volume and 
boron concentration requirements and Refueling Water Storage Tank 
(RWST) boron concentration requirements that are necessary to 
accommodate expected changes in the nuclear fuel (e.g., higher 
enrichment) that are associated with the planned power uprate. 
Additionally, the change would allow Ginna to utilize analysis 
methodologies that have been previously approved for use at 
Westinghouse nuclear plants. The changes to the TS are necessary to 
ensure the acceptability of these systems to perform their intended 
function in the event of an accident.
    The proposed changes do not adversely affect accident initiators 
or precursors nor significantly alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes do not 
adversely 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. Further, the proposed changes do not increase 
the types and amounts of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposures. The proposed changes cannot 
affect the probability of an accident occurring since they reflect a 
necessary change in plant design consistent with current design 
which is not an accident initiator. The proposed changes cannot 
increase the consequences of postulated accidents since they reflect 
a change in plant design that will continue to mitigate the effects 
of potential accidents. Therefore, the changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes include revising accumulator volume and 
boron concentration requirements and RWST boron concentration 
requirements that are necessary to accommodate expected changes in 
the nuclear fuel (e.g., higher enrichment) that are associated with 
the planned power uprate. Additionally, the change would allow Ginna 
to utilize analysis methodologies that have been previously approved 
for use at Westinghouse nuclear plants. The changes to the TS are 
necessary to ensure the acceptability of these systems to perform 
their intended function in the event of an accident.
    The proposed changes involve changes to accumulator volume and 
boron concentration requirements and RWST boron concentration 
requirements to ensure the continued acceptability of LOCA and post 
LOCA analysis results. The changes to the Technical Specifications 
(TS) are necessary to properly accommodate the changes in plant 
design. The changes ensure applicable acceptance criteria will 
continue to be met. The changes do not involve a significant change 
in the methods governing normal plant operation. The proposed TS 
changes do not create the possibility of a new or different [kind] 
of accident from those previously evaluated since they reflect a 
change that will ensure the accumulators and RWST will continue to 
perform their intended function in the event of an accident.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes include revising accumulator volume and 
boron concentration requirements and RWST boron concentration 
requirements that are necessary to accommodate expected changes in 
the nuclear fuel (e.g., higher enrichment) that are associated with 
the planned power uprate. Additionally, the change would allow Ginna 
to utilize analysis methodologies that have been previously approved 
for use at Westinghouse nuclear plants. The changes to the TS are 
necessary to ensure the acceptability of these systems to perform 
their intended function in the event of an accident.
    The level of safety of facility operation is not significantly 
affected by the proposed changes since there is no change in the 
intent of the TS requirements of assuring proper plant response in 
the event of an accident. The response of the plant systems to 
accidents and transients reported in the Updated Final Safety 
Analysis Report (UFSAR) is not adversely affected by this

[[Page 33220]]

change. Therefore, the capability to satisfy accident analysis 
acceptance criteria is not adversely affected. The proposed TS 
change cannot involve a significant reduction in [a] margin of 
safety since it is based upon changes that will maintain a 
substantial margin of safety with respect to accumulators and RWST 
functions. Therefore, the changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: April 29, 2005.
    Description of amendment request: The amendment would revise 
Technical Specifications (TSs) to allow the use of Relaxed Axial Offset 
Control (RAOC) methodology in reducing operator action required to 
maintain conformance with power distribution control TS and increasing 
the ability to return to power after a plant trip or transient while 
still maintaining margin to safety limits under all operating 
conditions.
    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 proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not initiate an accident. Evaluations 
and analyses of accidents, which are potentially affected by the 
parameters and assumptions, associated with the RAOC and 
FQ(Z) methodologies have shown that design standards and 
applicable safety criteria will continue to be met. The 
consideration of these changes does not result in a situation where 
the design, material, or construction standards that were applicable 
prior to the change are altered. Therefore, the proposed changes 
will not result in any additional challenges to plant equipment that 
could increase the probability of any previously evaluated accident.
    The proposed changes associated with the RAOC and 
FQ(Z) methodologies do not affect plant systems such that 
their function in the control of radiological consequences is 
adversely affected. The actual plant configurations, performance of 
systems, or initiating event mechanisms are not being changed as a 
result of the proposed changes. The design standards and applicable 
safety criteria limits will continue to be met; therefore, fission 
barrier integrity is not challenged. The proposed changes associated 
with the RAOC and FQ(Z) methodologies have been shown not 
to adversely affect the plant response to postulated accident 
scenarios. The proposed changes will therefore not affect the 
mitigation of the radiological consequences of any accident 
described in the Updated Final Safety Analysis Report (UFSAR).
    Therefore, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed changes do not challenge the performance or integrity 
of any safety-related system. The possibility for a new or different 
type of accident from any accident previously evaluated is not 
created since the proposed changes do not result in a change to the 
design basis of any plant structure, system or component. Evaluation 
of the effects of the proposed changes has shown that design 
standards and applicable safety criteria continue to be met.
    Equipment important to safety will continue to operate as 
designed and component integrity will not be challenged. The 
proposed changes do not result in any event previously deemed 
incredible being made credible. The proposed changes will not result 
in conditions that are more adverse and will not result in any 
increase in the challenges to safety systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes will not involve a significant reduction in 
a margin of safety.
    The proposed changes will assure continued compliance within the 
acceptance limits previously reviewed and approved by the NRC for 
RAOC and FQ(Z) methodologies. The appropriate acceptance 
criteria for the various analyses and evaluations will continue to 
be met.
    The projected impact associated with the implementation of RAOC 
on peak cladding temperature (PCT) has been incorporated into the 
LOCA [loss-of-coolant accident] analyses for the planned extended 
power uprate. It has [been] determined that implementation of RAOC 
at the extended power uprate power level does not result in a 
significant reduction in a margin of safety. The analysis performed 
for EPU [extended power uprate] bounds operation at the current 
power level.
    Therefore, the proposed changes do not involve a significant 
reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: May 5, 2005.
    Brief description of amendment request: The proposed amendment 
would change the Technical Specifications to modify the auxiliary 
feedwater (AFW) pump suction protection requirements and change the 
design basis as described in the Updated Safety Analysis Report to 
revise the functionality of the discharge pressure switches to provide 
pump runout protection, which requires operator actions to restore the 
AFW pumps for specific post-accident recovery activities.

[[Page 33221]]

    Date of publication of individual notice in Federal Register: May 
13, 2005 (70 FR 25619).
    Expiration date of individual notice: June 13, 2005.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: April 27, 2005, as supplemented May 4, 
2005.
    Description of amendment request: The proposed amendment would 
revise the SSES 1 and 2, Technical Specification 3.8.4, ``DC Sources-
Operating,'' to address new required actions for the condition in which 
a 125 volt direct current (VDC) charger is taken out of service for the 
purposes of a special inspection and related activities. The proposed 
changes would be in effect until the special inspection and related 
activities are completed on each of the 125 VDC Class 1E battery 
chargers but no later than 60 days following the issuance of the Unit 1 
and 2 amendments. Specifically, required Action A.2.1 would require 
that surveillance requirement 3.8.6.1 be performed within 2 hours and 
once-per-12 hours thereafter; and, required Action A.2.2 would restrict 
the restoration time for the inoperable electrical power subsystem to 
36 hours.
    Date of publication of individual notice in Federal Register: May 
12, 2005 (70 FR 25122).
    Expiration date of individual notice: Comments, May 27, 2005; 
Hearing, July 11, 2005.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating 
License, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing in connection with these actions was 
published in the Federal Register as indicated.

    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: October 21, 2004, as 
supplemented January 4, 2005.
    Brief description of amendment: The amendment deleted the Technical 
Specification (TS) requirements to submit monthly operating reports and 
annual occupational radiation exposure reports. The change is 
consistent with Revision 1 of NRC-approved Industry/Technical 
Specifications Task Force (TSTF) Standard TS Change Traveler, TSTF-369, 
``Removal of Monthly Operating Report and Occupational Radiation 
Exposure Report.'' This TS improvement was announced in the Federal 
Register (69 FR 35067) on June 23, 2004, as part of the Consolidated 
Line Item Improvement Process (CLIIP).
    Date of issuance: May 20, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 165.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19114).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 20, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: February 14, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specification Surveillance Requirement 3.3.7.1 to extend the 
frequency of the channel functional test for the Engineered Safeguards 
Protective System digital actuation logic channels from once every 31 
days to once every 92 days.
    Date of Issuance: May 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 345, 347 and 346.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12745).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 20, 2004, as supplemented by 
letter dated April 12, 2005.
    Brief description of amendment: The amendment deletes TS 6.6.1, 
``Occupational Radiation Exposure Report'' and TS 6.6.4, ``Monthly 
Operating Reports,'' as described in the Notice of Availability 
published in the Federal Register on June 23, 2004 (69 FR 35067).
    Date of issuance: May 13, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.

[[Page 33222]]

    Amendment No.: 259.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2890). The supplement dated April 12, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 13, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 22, 2004.
    Brief description of amendment: The requested change deletes 
Technical Specification (TS) 6.9.1.5, ``Occupational Radiation Exposure 
Report,'' and 6.9.1.6, ``Monthly Operating Reports,'' as described in 
the Notice of Availability published in the Federal Register on June 
23, 2004 (69 FR 35067).
    Date of issuance: May 25, 2005.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 202.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12746).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 25, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 27, 2005, as supplemented by 
letter dated May 12, 2005.
    Brief description of amendment: The amendment removed the license 
condition on instrument uncertainty that was imposed on the Waterford 3 
license with the issuance of License Amendment 199 for the extended 
power uprate.
    Date of issuance: May 23, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 201.
    Facility Operating License No. NPF-38: The amendment revised the 
Operating License.
    Date of initial notice in Federal Register: May 5, 2005 (70 FR 
23892). The May 12, 2005, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 23, 2005.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: September 15, 2004.
    Brief description of amendments: The amendments deleted the 
Technical Specification (TS) requirements related to hydrogen 
recombiners. The TS changes support implementation of the revisions to 
Title 10 of the Code of Federal Regulations (10 CFR) section 50.44, 
``Standards for Combustible Gas Control System in Light-Water-Cooled 
Power Reactors,'' that became effective on October 16, 2003. The 
changes are consistent with Revision 1 of the NRC-approved Industry/
Technical Specifications Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen 
Recombiners and Change to Hydrogen and Oxygen Monitors.''
    Date of issuance: May 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 137, 137, 143, 143.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5243).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2005.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: June 15, 2004, as supplemented 
January 12, 2005.
    Brief description of amendments: These amendments changed 
Surveillance Requirement (SR) 3.8.1.3, monthly diesel surveillance 
test; SR 3.8.1.10, diesel full load rejection test; SR 3.8.1.14.3.b, 
diesel 24-hour run test; and, SR 3.8.1.15, diesel hot restart test, to 
permit these tests to be run at a higher load up to 2800 kW.
    Date of issuance: May 20, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendments Nos.: 253 and 256.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Technical Specifications.
    Date of initial notice in  Federal Register: July 20, 2004, (69 FR 
43461). The January 12, 2005, supplement provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on July 20, 2004 (69 FR 43461).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 20, 2005.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 21, 2004, as supplemented by letters 
dated September 16, and December 14, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specification Bases Section to allow the containment spray pumps to be 
secured during a loss-of-coolant accident, when certain conditions are 
met, to minimize the potential for containment sump clogging.
    Date of issuance: May 20, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 120 days of issuance.
    Amendment No.: 235.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications Bases.
    Date of initial notice in  Federal Register: June 22, 2004 (69 FR 
34703). The September 16, and December 14,

[[Page 33223]]

2004, 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 staff's original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated May 20, 2005.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: May 21, 2004.
    Brief description of amendment: The amendment revises Technical 
Specifications related to the reactor coolant pump flywheel inspection 
program by relocating the requirements from the limiting conditions for 
operation to the administrative controls section and increasing the 
inspection interval to 20 years.
    Date of issuance: May 9, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 172.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Technical Specifications.
    Date of initial notice in  Federal Register: March 1, 2005 (70 FR 
9995).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 9, 2005.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: October 21, 2004, as supplemented 
December 13 and 22, 2004, and February 23 and March 1, 2005.
    Brief description of amendments: Conforming license amendments to 
remove AEP Texas Central Company as an ``Owner'' in the facility 
operating licenses.
    Date of issuance: May 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 365 days of issuance.
    Amendment Nos.: Unit 1-172; Unit 2-160
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the licenses.
    Date of initial notice in Federal Register: December 14, 2004 (69 
FR 76019). The supplements dated December 13 and 22, 2004, and February 
23 and March 1, 2005, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2005.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License, 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

[[Page 33224]]

System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the PDR Reference staff at 1 (800) 
397-4209, (301) 415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
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.\1\ 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 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.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
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    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    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, [email protected]; 
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 
[email protected]. 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 or 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).

[[Page 33225]]

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of amendment request: April 26, 2005, as supplemented on April 
29 and on May 3, 2005.
    Description of amendment request: Revises the Completion Time for 
the Action associated with an inoperable low pressure Emergency Core 
Cooling System injection/spray system to 14 days on a one-time basis.
    Date of issuance: May 9, 2005.
    Effective date: As of date of issuance and shall be implemented 
within 7 days.
    Amendment No.: 294.
    Facility Operating License No. DPR-52: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of NSHC determination 
are contained in a Safety Evaluation dated May 9, 2005.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

    Dated in Rockville, Maryland, this 27th day of May 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. E5-2848 Filed 6-6-05; 8:45 am]
BILLING CODE 7590-01-P