[Federal Register Volume 70, Number 99 (Tuesday, May 24, 2005)]
[Notices]
[Pages 29785-29808]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-10063]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 29, 2005 through May 12, 2005. The 
last biweekly notice was published on May 10, 2005 (70 FR 24645).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, 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

[[Page 29786]]

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://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].

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

    Date of amendment request: January 21, 2005.
    Description of amendment request: The proposed amendment would 
implement the Alternative Source Term (AST) for the analysis of the 
radiological consequences of a design-basis Loss-of-Coolant Accident 
(LOCA). There are no changes proposed to the Operating License or 
Technical Specifications.
    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

    Revision of the LOCA analysis to the Alternative Source Term 
methodology does not affect the design or operation of HBRSEP [H. B. 
Robinson Steam Electric Plant], Unit No. 2. Rather, once the 
occurrence of an accident has been postulated, the new source term 
is an input to evaluate the consequences of the postulated accident. 
The implementation of the Alternative Source Term has been evaluated 
in revisions to the LOCA dose analysis at HBRSEP, Unit No. 2. Based 
on the results of this analysis, it has been demonstrated that the 
dose consequences are within the regulatory guidance provided by the 
NRC. This guidance is presented in 10 CFR 50.67 and Regulatory Guide 
1.183.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

2. The Proposed Change Does Not Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated

    The proposed change does not affect plant structures, systems, 
or components. The proposed change is to an evaluation methodology 
and does not initiate design basis accidents.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

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

    The proposed change is associated with the implementation of a 
new licensing basis for HBRSEP, Unit No. 2. The new licensing basis 
implements an Alternative Source Term in accordance with 10 CFR 
50.67 and the associated Regulatory Guide 1.183. The results of the 
revised limiting design basis analysis are subject to revised 
acceptance criteria. The analysis has been performed using 
conservative methodologies in accordance with regulatory guidance or 
other methodologies approved by the NRC in prior plant-specific 
license amendments. The dose consequences are within the acceptance 
criteria found in the regulatory guidance associated with 
Alternative Source Terms.
    The proposed change continues to ensure that doses at the 
exclusion area and low population zone boundaries, as well as the 
control room, are within the corresponding regulatory limits. 
Specifically, the margin of safety for the radiological consequences 
of these accidents is considered to be that provided by meeting the 
applicable regulatory limits.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

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

[[Page 29787]]

    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: February 14, 2005.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements (SRs) for the station batteries as 
specified in Technical Specification (TS) SR 3.8.4.5, the battery 
service test, and TS SR 3.8.4.6, the battery performance test.
    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?
    No. The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed surveillance changes will continue to ensure 
that the DC system is tested in a manner that will verify 
operability. Performance of the required system surveillances, in 
conjunction with the applicable operational and design requirements 
for the DC system, provide assurance that the system will be capable 
of performing the required design functions for accident mitigation 
and also that the system will perform in accordance with the 
functional requirements for the system as described in the Updated 
Final Safety Analysis Report for HBRSEP [H. B. Robinson Steam 
Electric Plant], Unit No. 2. This ensures that the rate of 
occurrence and consequences of analyzed accidents will not change. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the Proposed Changes Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated?
    No. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated. The 
proposed surveillance requirement changes will continue to ensure 
that the DC system is tested in a manner that will verify 
operability. No physical changes to the HBRSEP, Unit No. 2, systems, 
structures, or components are being implemented. There are no new or 
different accident initiators or sequences being created by the 
proposed Technical Specifications changes. Therefore, these changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the Proposed Changes Involve a Significant Reduction in 
the Margin of Safety?
    No. The proposed changes do not involve a significant reduction 
in the margin of safety. The proposed DC system surveillance 
requirement changes provide appropriate and applicable surveillances 
for the DC system. The proposed changes to surveillance requirements 
for the DC system will continue to ensure system operability. 
Therefore, these changes do not affect any margin of safety for 
HBRSEP, Unit No. 2.

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

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

    Date of amendment request: March 3, 2005.
    Description of amendment request: The proposed amendment would 
revise the requirements of Technical Specification (TS) 5.6.5, ``Core 
Operating Limits Report (COLR).'' Specifically, the proposed change 
would add topical report EMF-2103(P)(A), ``Realistic Large Break LOCA 
[loss-of-coolant accident] Methodology for Pressurized Water 
Reactors,'' to the list of documents specified in TS 5.6.5. TS 5.6.5 
lists the approved methodologies that can be used to determine the core 
operating limits.
    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 methodology will be reviewed and approved by the 
NRC prior to its use for HBRSEP [H. B. Robinson Steam Electric 
Plant], Unit No. 2. Analyzed events are assumed to be initiated by 
the failure of plant structures, systems, or components. The 
determination of core operating limits in accordance with this new 
methodology will meet the limitations specified in the NRC safety 
evaluation of the new methodology. The topical report associated 
with the new methodology demonstrates that the integrity of the fuel 
will be maintained and that design requirements will continue to be 
met. The proposed change does not involve physical changes to any 
plant structure, system, or component. Therefore, the probability of 
occurrence for a previously analyzed accident is not significantly 
increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. The proposed methodology continues to meet 
applicable design and safety analyses acceptance criteria. The 
proposed change does not affect the performance of any equipment 
used to mitigate the consequences of an analyzed accident. As a 
result, no analysis assumptions are violated and there are no 
adverse effects on the factors that contribute to offsite or onsite 
dose as the result of an accident. The proposed change does not 
affect setpoints that initiate protective or mitigative actions. The 
proposed change ensures that plant structures, systems, or 
components are maintained consistent with the safety analysis and 
licensing bases. Based on this evaluation, there is no significant 
increase in the consequences of a previously analyzed event. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures, or components, other than allowing for 
fuel design in accordance with NRC approved methodologies. The 
proposed methodology continues to meet applicable criteria for Large 
Break Loss of Coolant Accident (LBLOCA) analysis. No new or 
different equipment is being installed. No installed equipment is 
being operated in a different manner. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints that initiate protective or mitigative actions. As a 
result, no new failure modes are being introduced. There are no 
changes in the methods governing normal plant operation, nor are the 
methods utilized to respond to plant transients altered. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety?
    The margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The proposed change in the methodology used for LBLOCA analyses does 
not impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. The 
proposed

[[Page 29788]]

change does not significantly impact any safety analysis assumptions 
or results. Therefore, the proposed change does not result in a 
significant reduction in the margin of safety.

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

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Power Station, Unit Nos. 1, 2, and 3, New London County, 
Connecticut

    Date of amendment request: December 21, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) requirements for annual Occupational 
Radiation Exposure Reports (all units), annual report regarding 
challenges to pressurizer relief and safety valves (Units 2 and 3), and 
Monthly Operating Reports (Units 2 and 3).
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
license amendment applications in the Federal Register on June 23, 2004 
(69 FR 35067). The licensee affirmed the applicability of the model 
NSHC determination in its application dated December 21, 2004.
    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 eliminates the TSs reporting requirements to 
provide a monthly operating letter report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the TS reporting 
requirement for an annual occupational radiation exposure report, 
which provides information beyond that specified in NRC regulations. 
The proposed change involves no changes to plant systems or accident 
analyses. As such, the change is administrative in nature and does 
not affect initiators of analyzed events or assumed mitigation of 
accidents or transients. 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 does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-336, Millstone 
Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: March 9, 2005.
    Description of amendment request: Current Technical Specifications 
(TSs) require that all operations involving a reduction in reactor 
coolant boron concentration or that involve positive reactivity changes 
be suspended under certain conditions. The requested changes modify the 
TSs to incorporate wording related to the reactor coolant system (RCS), 
electrical power systems, and refueling operations to provide 
operational flexibility during mode changes or addition of coolant 
during shutdown operations. Additionally, changes are to be made to the 
TS bases, as appropriate.
    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:

Criterion 1: Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change does not in any way alter the SDM [shutdown 
margin] or refueling boron concentration. It limits introduction of 
coolant into the RCS of reactivity more positive than that necessary 
to meet the required SDM or refueling boron concentration. This 
proposed change does not affect the input or assumptions for any 
accidents previously evaluated nor does it affect initiation of an 
accident. Based on this discussion, the proposed amendment does not 
increase the probability or consequence of an accident previously 
evaluated.

Criterion 2: Does the proposed amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?

    Response: No.
    The proposed change allows introduction of coolant into the RCS 
with different temperature or lower boron concentration, however, 
the required boron concentration or SDM is maintained. The proposed 
amendment does not introduce failure modes, accident initiators, or 
malfunctions that would cause a new or different kind of accident. 
No plant modifications are associated with the change. Therefore, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?

    Response: No.
    The proposed change provides the flexibility necessary for 
continued safe reactor operations while limiting any potential for 
excess positive reactivity additions. [The] SDM and required boron 
concentration are not affected. Therefore, based on the above, the 
proposed amendment 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell J. Roberts.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: December 23, 2004.
    Description of amendment request: The requested amendment would

[[Page 29789]]

relocate certain Technical Specifications regarding refueling 
operations to the Technical Requirements Manual (TRM).
    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:

Criterion 1: Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously evaluated?

    Response: No.
    The communications equipment, refueling machine, and spent fuel 
pool crane are not designed to perform accident mitigation 
functions. The proposed change to relocate selected refueling 
specifications does not modify any plant equipment and does not 
impact any failure modes that could lead to an accident. Relocating 
the specifications to the TRM where changes would be controlled 
under the 10 CFR 50.59 process does not change the ability of the 
communications or refueling equipment to function as expected. 
Additionally, these specifications have no affect on the consequence 
of any analyzed accident since the equipment is not related to 
accident mitigation. Based on this discussion, the proposed 
amendment does not increase the probability or consequences of an 
accident previously evaluated.

Criterion 2: Does the proposed amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?

    Response: No.
    The proposed change[s] do[es] not modify any plant equipment and 
there is no impact on the capability of the existing equipment to 
perform their intended functions to move fuel safely or conduct 
refueling operations while in contact with the control room. No 
system setpoints are being modified and no changes are being made to 
the method in which refueling operations are conducted. No changes 
to the heavy loads program are being proposed by this change. No new 
failure modes are introduced by the proposed changes. The proposed 
amendment does not introduce accident initiators or malfunctions 
that would cause a new or different kind of accident. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?

    Response: No.
    The relocation of Technical Specification 3/4.9.5, ``Refueling 
Operations, Communications,'' to the TRM does not imply any 
reduction in its importance in [e]nsuring communication between the 
control room and the refueling station. The proposed change will not 
alter the requirement on communication between the control room and 
the refueling station, it will not alter any of the assumptions used 
in the fuel handling accident analysis, nor will it cause any safety 
system parameters to exceed their acceptance limit. The relocation 
of Technical Specification 3/4.9.6, ``Refueling Machine'' to the TRM 
does not alter the requirement for the lifting device on the 
refueling machine to have adequate capacity or for the interlocks to 
be demonstrated operable prior to fuel movement. The assumptions 
used in the accident analysis are not impacted by this change and no 
impact to any safety system parameters will result. The relocation 
of Technical Specification 3/4.9.7, ``Crane Travel--Spent Fuel 
Storage Areas,'' to the TRM will not alter the requirement that the 
crane interlocks and/or physical stops are operable, nor will it 
alter any of the assumptions used in the fuel handling accident 
analysis. Heavy load lifts are administratively controlled by a safe 
load path and crane interlocks. The proposed change[s] do[es] not 
modify any heavy load path criteria. Administrative changes 
associated with the proposed revision such as relocation of 
associated Technical Specification Bases to the TRM will not have an 
impact on any established safety margins.
    The proposed change[s] do[es] not affect any of the assumptions 
used in the accident analysis, nor do they affect any operability 
requirements for equipment important to plant safety. Therefore, the 
proposed change[s] will not result in a significant reduction in the 
margin of safety as defined in the Bases for Technical 
Specifications covered in this License Amendment Request.

    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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell J. Roberts.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: November 25, 2002, as supplemented by 
letters dated November 13, and December 16, 2003, September 22, 2004, 
and April 6, 2005.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) for the Ventilation Filter Testing 
Program (VFTP), Annulus Ventilation System (AVS), Auxiliary Building 
Filtered Ventilation Exhaust System (ABFVES), Fuel Handling Ventilation 
Exhaust System (FHVES), and Control Room Area Ventilation System 
(CRAVS), and containment penetrations.
    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:

First Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated? No.
    This license amendment request proposes amendments to the system 
TS and/or Bases and/or VFTP TS requirements for the AVS, ABFVES, 
FHVES, and CRAVS. It also proposes amendments to the TS and Bases 
for Containment Penetrations. The AVS is in standby during normal 
plant operations and operates only following a Safety Injection 
signal or during a test. It is not an accident initiator. The ABFVES 
is in operation during normal plant operations. However, the ABFVES 
is not used in direct support of any phase of power generation or 
conversion or transmission, shutdown cooling, fuel handling 
operations, or processing of radioactive fluids. Therefore, it is 
not an accident initiator. The FHVES is utilized to support fuel 
handling operations when moving recently irradiated fuel. It is not 
an accident initiator. The CRAVS operates during normal plant 
operations. However, it is not an accident initiator (the CRAVS 
being defined so as to exclude equipment that maintains an 
appropriately low temperature in the control room). The status of 
containment penetrations is required to be controlled so as to 
minimize the consequences of a fuel handling accident or a weir gate 
drop accident. The containment penetrations by themselves are not 
accident initiators. No accident initiators are associated with the 
changes proposed in this license amendment request. For these 
reasons, operation of the facility in accordance with this proposed 
amendment does not involve a significant increase in the probability 
of any accident previously evaluated.
    In support of the proposed amendment, an analysis has been 
performed to determine the radiological consequences of the design 
basis [Loss of Coolant Accident] LOCA at Catawba Nuclear Station. 
The analysis made use of the Alternative Source Term (AST) 
methodology and in general conformed to the regulatory positions of 
Regulatory Guide 1.183 and the draft regulatory positions of DG-
1111. Total Effective Dose Equivalent (TEDE) radiation doses at the 
Exclusion Area Boundary (EAB), boundary of the Low Population Zone 
(LPZ), and to the control room operators were calculated and found 
to be acceptable. TEDEs were calculated for a design basis LOCA 
postulated for a Catawba nuclear unit operating with all low 
enriched uranium (LEU) fuel and with 4 mixed oxide (MOX) lead fuel 
assemblies (LFAs). It was found that insertion of 4 MOX LFAs did not 
produce a significant increase in the TEDEs for a design basis LOCA.
* * * * *
    The new value for the control room TEDE radiation dose is higher 
than the TEDE radiation dose equivalent to the radiation

[[Page 29790]]

doses currently reported in the UFSAR. However, the limiting control 
room TEDE radiation dose reported in this submittal is lower than 
the acceptance criterion * * * The new LPZ TEDE radiation dose is 
higher than the equivalent TEDE radiation dose currently 
represented. On the other hand, the margin to the acceptance 
criterion is [large] * * *. The TEDE radiation doses newly computed 
at the EAB for the design basis LOCA are lower than the 
corresponding equivalent EAB TEDE radiation dose currently 
represented in the UFSAR. The margin in the EAB TEDE radiation dose 
to the guideline value is [also large]. * * * In all cases, there is 
significant margin between the newly calculated post-LOCA TEDE 
radiation doses and the corresponding regulatory guideline values. 
In the sense that the margins to the germane regulatory guideline 
values are still large, the new values of TEDE radiation doses are 
comparable to the equivalent TEDE associated with the post-LOCA 
radiation doses currently listed in the UFSAR. Furthermore, these 
margins for the design basis LOCA do not significantly decrease with 
insertion of the 4 MOX LFAs. Therefore, the proposed amendment is 
determined to not result in a significant increase in accident 
consequences.
    AST analyses also were completed for the design basis locked 
rotor accident (LRA) and rod ejection accident (REA). Again, these 
design basis accidents were postulated to occur at a Catawba nuclear 
unit operating with either an all LEU core or with 4 MOX LFAs. The 
TEDEs following these design basis accidents were compared to the 
equivalent TEDEs associated with the current license basis analyses. 
The equivalent TEDEs were computed from the post-accident whole body 
and thyroid radiation doses using the method prescribed in 
Regulatory Guide 1.183, as noted above. TEDEs only at offsite 
locations were compared as post-accident control room radiation 
doses are not reported for these design basis accidents in the 
Catawba UFSAR.
* * * * * * *
    For the EAB, LPZ, and control room, the post-LRA TEDEs are seen 
to increase from the values equivalent to the radiation doses from 
the current license basis analyses. (This is attributed primarily to 
the increase in assumed fraction of the fuel pins with clad failure 
following a design basis LRA at Unit 2. * * *) However, the margins 
to the acceptance criteria of 2.5 Rem at the offsite locations and 5 
Rem in the control room are still significant.
* * * * * * *
    For the EAB, LPZ, and control room, the post-REA TEDEs are seen 
to increase from the values equivalent to the radiation doses from 
the current license basis analyses, as they did for the design basis 
LRA. (This is attributed to a number of reasons. These include 
increase in the fraction of gap activity released to containment, 
inclusion of limiting radial peaking in the source term, and 
inclusion of alkali metals.) However, the margins to the acceptance 
criteria of 6.3 Rem at the offsite locations and 5 Rem in the 
control room are still significant * * *.
    The changes proposed to the TS for Containment Penetrations are 
editorial in nature and will have no effect upon accident 
consequences.
    The changes proposed to the VFTP TS for the AVS, ABFVES, and 
FHVES will not result in a significant increase in any accident 
consequences. The changes to make the penetration values for Unit 2 
consistent with Unit 1 for the AVS, ABFVES, and FHVES are acceptable 
because the appropriate safety factors as delineated in the 
applicable regulatory guideline documents are still maintained. The 
change to the flowrate specified for the ABFVES is consistent with 
the design basis operation of this system. Also, the editorial 
changes proposed to the VFTP TS will have no impact on any 
accidents.
    Operation of the facility in accordance with the proposed 
amendment does not involve a significant increase in the 
consequences of an accident previously evaluated.

Second Standard

    Does operation of the facility in accordance with the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated? No.
    This proposed amendment does not involve addition, removal, or 
modification of any plant system, structure, or component. These 
changes will not affect the operation of any plant system, 
structure, or components as directed in plant procedures.
    The analysis performed in support of this license amendment 
request, together with the analyses of the design basis fuel 
handling accident and weir gate drop reported in previously 
submitted and NRC approved license amendment requests, includes full 
scope implementation of AST methodology. This analysis does not 
represent any change in the post-accident operation of any plant 
system, structure, or component.
    Operation of the facility in accordance with this amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

Third Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety? 
No.
    Margin of safety is related to confidence in the ability of 
fission product barriers to perform their design functions following 
any of their design basis accidents. These barriers include the fuel 
cladding, the Reactor Coolant System, and the containment. The 
performance of these barriers either during normal plant operations 
or following an accident will not be affected by the changes 
associated with the license amendment request.
    The AVS is associated with the containment fission product 
barrier. Its post-accident operation will not be affected by 
implementation of the amendment to its TS. The operation of the 
ABFVES either during normal plant operations or following an 
accident will not be affected by implementation of the amendment to 
its TS. The operation of the FHVES either during normal plant 
operations or following an accident will not be affected by 
implementation of the amendment to its TS. The operation of the 
CRAVS either during normal plant operations or following an accident 
will not be adversely affected by the proposed changes to its TS 
Bases. The operation of Containment Penetrations following an 
accident will not be adversely affected by the proposed change to 
its TS.
    As noted, an analysis of radiological consequences of the design 
basis LOCA at Catawba Nuclear Station has been performed in support 
of this license amendment request. The design basis LOCA scenarios 
were selected based on extensive evaluations of Catawba, its design 
basis, and its anticipated response to a design basis LOCA. Credit 
was taken only for safety related systems, structures, and 
components in simulating the mitigation of radiological consequences 
of the LOCA. Limiting values were taken for performance 
characteristics of the Class 1E systems modeled in the analysis. The 
radiological consequences (TEDE radiation doses at the EAB, LPZ, and 
in the control room) are within the regulatory guideline values with 
significant margin.
    The changes proposed to the VFTP TS for the AVS, ABFVES, and 
FHVES will not result in a significant reduction in the margin of 
safety. These changes are supported by regulatory guidance 
documents, and are consistent with existing system operation. Also, 
the editorial changes proposed to the VFTP TS will not have any 
impact on safety.
    Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas
    Date of amendment request: September 30, 2004, as supplemented by 
letter dated April 26, 2005.
    Description of amendment request: The proposed amendment would 
change the existing steam generator (SG) tube surveillance program to 
be consistent with that being proposed by the Technical Specifications 
Task Force (TSTF) in TSTF-449. These proposed changes would revise 
Technical Specification (TS) 1.1 on definitions, TS 3.4.13 on reactor 
coolant system

[[Page 29791]]

operational leakage, TS 5.5.9 on SG program, and TS 5.6.7 on SG tube 
inspection reports, and add a new TS 3.4.16 on SG tube integrity. Also, 
as a result of the licensee replacing the SGs with SGs having a new 
Alloy 690 thermally treated tubing design, the TSs would be revised to 
reflect this replacement. The September 30, 2004, application was 
noticed in the Federal Register on November 9, 2004 (69 FR 64987).
    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 requires a Steam Generator Program that 
includes performance criteria that will provide reasonable assurance 
that the steam generator (SG) tubing will retain integrity over the 
full range of design basis operating conditions (including startup, 
power operation, hot standby, cooldown, anticipated transients and 
postulated accidents). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE. These criteria assure that the probability of an accident 
will not be increased.
    The primary to secondary accident induced leakage rate for any 
design basis accidents, other than an SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. [The primary to secondary accident induced leakage rate is 
relatively inconsequential for the SG tube rupture analysis.] The 
operational LEAKAGE performance criterion meets current NRC 
regulations and NEI [Nuclear Energy Institute] 97-06 criteria for 
reactor coolant system (RCS) operational primary to secondary 
LEAKAGE through any one SG of 150 gallons per day. These criteria 
assure that accident doses will stay within regulatory and licensing 
basis limits.
    Therefore, the proposed change does not affect the probability 
or consequences of any ANO-1 [Arkansas Nuclear One, Unit 1] analyzed 
accidents.
    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 performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed Steam Generator 
Program will not introduce any adverse changes to the plant design 
basis or postulated accidents resulting from potential tube 
degradation. The proposed change does not affect the design of the 
SGs, their method of operation, or primary or secondary coolant 
chemistry controls. The proposed change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the Steam Generator Program to manage SG 
tube inspection, assessment, repair, and plugging. The requirements 
established by the Steam Generator Program are consistent with those 
in the applicable design codes and standards and are an improvement 
over the requirements in the current technical specifications.
    Therefore, the margin of safety is not changed by the proposed 
change to the ANO-1 TSs.

    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: March 30, 2005.
    Description of amendment request: The proposed amendment adopts the 
following Nuclear Regulatory Commission (NRC) approved Technical 
Specification Task Force (TSTF) changes that affect the Boiling Water 
Reactor (BWR)/6 Improved Standard Technical Specifications:

--------------------------------------------------------------------------------------------------------------------------------------------------------
            TSTF No.                                 Description                            TS section affected                  Type of change
--------------------------------------------------------------------------------------------------------------------------------------------------------
046, Rev. 1.....................  Clarify the Containment Isolation Valve            SR 3.6.1.3.4....................  Administrative.
                                   Surveillance Requirement (SR) to apply only to    SR 3.6.4.2.2....................
                                   automatic isolation valves.                       SR 3.6.5.3.3....................
222, Rev. 1.....................  Control Rod Scram Time Testing...................  SR 3.1.4.1......................  Administrative.
                                                                                     SR 3.1.4.4......................
264, Rev........................  Delete flux monitors specific overlap SRs........  SR 3.3.1.1.5....................  Less Restrictive.
                                                                                     SR 3.3.1.1.6....................
                                                                                     Table 3.3.1.1-1.................
275, Rev. 0.....................  Clarify requirements for Diesel Generator (DG)     Table 3.3.5.1-1, Footnote (a)...  Administrative.
                                   start signal on Reactor Pressure Vessel (RPV)
                                   level--low, low, low during RPV cavity flood-up.
276, Rev. 2.....................  Revise DG full load rejection test...............  SR 3.8.1.9......................  Less Restrictive.
                                                                                     SR 3.8.1.10.....................
                                                                                     SR 3.8.1.14.....................
300, Rev. 0.....................  Eliminate DG loss of coolant accident-Start SRs    SR 3.8.2.1......................  Less Restrictive.
                                   while in shutdown when emergency core cooling
                                   system is not required.
322, Rev. 2.....................  Secondary Containment Integrity SRs..............  SR 3.6.4.1.3....................  Administrative.
                                                                                     SR 3.6.4.1.4....................
400, Rev. 1.....................  Clarification of SR on bypass of DG automatic      SR 3.8.1.13.....................  Administrative.
                                   trips.
416, Rev. 0.....................  SR 3.5.1.2 Notation..............................  LCO 3.5.1.......................  Administrative.
                                                                                     SR 3.5.1.2......................
                                                                                     LCO 3.5.2.......................
                                                                                     SR 3.5.2.4......................
--------------------------------------------------------------------------------------------------------------------------------------------------------


[[Page 29792]]

    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 changes to the TS [Technical Specifications] 
involve both administrative and less restrictive changes. The 
administrative changes involve wording changes that clarify 
requirements without changing the original intent. As such, these 
types of changes do not affect initiators of analyzed events and do 
not affect the mitigation of any accidents or transients.
    The less restrictive changes involve modifications to 
Surveillance Requirements. The modified Surveillance Requirements do 
not cause the plant to be operated in a new or different manner and 
the required equipment continues to be tested in a manner and at a 
frequency necessary to provide confidence that the equipment can 
perform its intended safety function. Consequently, no initiators to 
accidents previously evaluated are affected and no mitigating 
equipment assumed in accidents previously evaluated is adversely 
affected.
    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 changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed), do 
not change the design function of any equipment, and do not change 
the methods of normal plant operation. Accordingly, the proposed 
changes do not create any new credible failure mechanisms, 
malfunctions, or accident initiators not previously considered in 
the GGNS [Grand Gulf Nuclear Station] design and licensing basis.
    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 changes have no affect on any safety analysis 
assumptions or methods of performing safety analyses. The changes do 
not adversely affect system OPERABILITY or design requirements and 
the equipment continues to be tested in a manner and at a frequency 
necessary to provide confidence that the equipment can perform its 
intended safety functions. 10 CFR 50.36 (c)(3) requires the TS to 
include Surveillance Requirements relating to test, calibration, or 
inspection to assure that the necessary quality of systems and 
components is maintained, that facility operation will be within 
safety limits, and that the limiting conditions for operation will 
be met. The GGNS TS Surveillance Requirements will continue to 
provide this assurance with the proposed adoption of the NRC 
approved TSTF changes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Allen G. Howe.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: December 14, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.G, ``Scram Discharge Volume 
[SDV],'' to allow vent or drain lines with inoperable valves to be 
isolated instead of requiring the valves to be restored to Operable 
status or to be 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 as a result of the Pilgrim Nuclear 
Power Station TS format) in its application dated December 14, 2004.
    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 vent or drain valves inoperable instead of requiring the 
valves to be restored to operable status or be in Hot Shutdown 
within 12 hours. With one SDV vent or drain valve inoperable in one 
or more lines, the isolation function would be maintained since the 
redundant valve in the affected line would perform its safety 
function of isolating the SDV. Following the completion of the 
required action, the isolation function is fulfilled since the 
associated line is isolated. The ability to vent and drain the 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell J. Roberts.

[[Page 29793]]

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: January 21, 2005.
    Description of amendment request: The proposed change permanently 
revises Isolation Condenser (IC) Technical Specifications (TS) Section 
3.5.3, ``IC System.'' Specifically, surveillance requirement SR 3.5.3.4 
is modified by the addition of a note which states the IC System heat 
removal capability surveillance is not required to be performed until 
12 hours after adequate reactor power is achieved to perform the test.
    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:

    According to 10 CFR 50.92, ``Issuance of amendment,'' paragraph 
(c), a proposed amendment to an operating license involves a no 
significant hazards consideration if 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;
    (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.
    In support of this determination, an evaluation of each of the 
three criteria set forth in 10 CFR 50.92 is provided below regarding 
the proposed license amendment.
    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The design function of the Isolation Condenser (IC) System is to 
provide reactor core cooling in the event that the reactor becomes 
isolated from the turbine and the main condenser by closure of the 
main steam isolation valves (MSIVs). Although the system is an 
Engineered Safety Feature System, no credit for IC System operation 
is taken in the accident analysis. The IC System is designed and 
installed to provide adequate core cooling, thereby mitigating the 
consequences of this reactor isolation transient (e. g., inadvertent 
closure of the MSIVs). This transient has been evaluated in the 
Updated Final Safety Analysis Report (UFSAR) as an event of moderate 
frequency. The IC system is designed to operate automatically or 
manually to perform its design function for reactor pressures 
greater than 150 psig. Since the IC System is not credited, this TS 
change does not impact any of the assumptions, inputs, or results of 
the UFSAR reactor isolation analysis.
    The addition of the note to the Technical Specifications 
surveillance requirement does not alter the IC System design 
function or the processes and parameters by which the system and its 
components perform its function. The addition of this note allows 
the plant to enter an operating mode necessary to allow performance 
of the heat removal capability surveillance. The purpose of this 
heat removal capability surveillance is to verify proper flow path 
and the ability to remove a design heat load. The proposed change 
does not alter the ability or methods used to verify flow path or 
heat removal capability. Nor does the change alter the acceptance 
criteria for satisfactory performance. Therefore, the change does 
not result in an increase in the consequences of a reactor isolation 
transient. Additionally, there are no IC System malfunctions or 
component failures that could initiate a reactor isolation 
transient. The proposed change does not alter the system or its 
operation and will not change the IC System's impact on initiating 
accidents or transients. Therefore, this change, and any associated 
impacts, will not increase the probability of the occurrence of an 
accident or transient.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The addition of the note to the Technical Specifications 
surveillance requirement does not alter the IC System design 
function or the processes and parameters by which the system and its 
components perform its function. The existing Technical 
Specification does not provide any limitations on when the IC System 
heat removal capability surveillance may be performed. Present plant 
procedures perform this surveillance at between 60% and 75% reactor 
power to ensure sufficient steam is available to simulate design 
heat loads. The addition of the note to the Technical Specification 
does not create any constraints on plant operating conditions 
associated with performance of the IC System heat removal capability 
surveillance. Operation of the IC System to perform the required 
surveillance in operating Modes 1, 2, or 3 has been previously 
evaluated and is presently allowed.
    The proposed change does not modify the procedural steps for 
performing the Technical Specification required surveillance. Nor 
does the change alter the methodology for evaluating acceptable 
performance. No physical or operational changes are made that could 
result in plant or system operation in conditions not previously 
evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Technical Specification surveillance requirement SR 3.5.3.4 
requires verification of the IC System's heat removal capability 
every 60 months. This surveillance ensures the proper system flow 
path and ability to remove decay heat following a reactor isolation. 
The methodology and acceptance criteria for this surveillance are 
not impacted by this change. Technical Specifications presently 
allow performance of this surveillance in Modes 1, 2, or 3 and plant 
procedures presently perform this surveillance in Mode 1. The 
surveillance is still required to demonstrate the IC System design 
basis capability of removing the design requirement of 252.5 x 
106 Btu/hr. Other IC System surveillance requirements are 
not directly or indirectly impacted by this change. Additionally, 
this amendment request results in no change to the system's 
actuation response, operation, or setpoints for performance.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: June 11, 2004.
    Brief description of amendment request: The proposed license 
amendment request would relocate surveillance test intervals of various 
Technical Specification (TS) surveillance requirements to a new program 
controlled in accordance with the requirements of 10 CFR 50.59. The 
proposed changes would add a new program, the Surveillance Frequency 
Control Program, to the Administrative Controls section of the TSs. The 
proposed amendment is a pilot submittal in support of the Boiling Water 
Reactor Owners' Group Risk-Informed Initiative 5b, ``Relocate 
Surveillance Test Intervals to Licensee Control.''
    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

[[Page 29794]]

consequences of an accident previously evaluated?
    Response: No. The proposed change involves the relocation of 
various surveillance test intervals from Technical Specifications 
(TS) to a licensee-controlled program and is administrative in 
nature. The proposed change does not involve the modification of any 
plant equipment or affect basic plant operation. The proposed change 
will have no impact on any safety related structures, systems or 
components. Surveillance test intervals are not assumed to be an 
initiator of any analyzed event, nor are they assumed in the 
mitigation of consequences of accidents. The surveillance 
requirements themselves will be maintained in TS[s] along with the 
applicable Limiting Conditions for Operation (LCOs) and Action 
statements. The surveillances performed at the intervals specified 
in the licensee-controlled program will assure that the affected 
system or component function is maintained, that the facility 
operation is within the Safety Limits, and that the LCOs are met.
    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 previously evaluated?
    Response: No. The proposed change does not involve any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function or is tested. 
As such, no new or different types of equipment will be installed, 
and the basic operation of installed equipment is unchanged. The 
methods governing plant operation and testing remain consistent with 
the safety analysis assumptions.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed change is administrative in nature, 
does not negate any existing requirement, and does not adversely 
affect existing plant safety margins or the reliability of the 
equipment assumed to operate in the safety analysis. As such, there 
are no changes being made to safety analysis assumptions, safety 
limits or safety system settings that would adversely affect plant 
safety as a result of the proposed change. Margins of safety are 
unaffected by relocation of the surveillance test intervals to a 
licensee-controlled program.
    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. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Darrell J. Roberts.

Exelon Generation Company, LLC, Docket No. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: July 22, 2004, as supplemented December 
3, 2004.
    Description of amendment request: The proposed amendment would 
modify the operability and surveillance requirements in Technical 
Specification 3/4.1.3, ``Control Rods.'' Specifically, the proposed 
changes would (1) exclude a fully inserted immovable control rod from 
the shutdown action statement, (2) eliminate consideration of control 
rod drive water pressure in the action statement, and (3) limit the 24-
hour exercise test of other control rods to a one-time occasion 
following detection of an immovable control rod.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The first proposed change would exclude fully 
inserted immovable control rods from consideration in the plant 
shutdown action statement. An inoperable control rod that has been 
fully inserted, and disarmed, has satisfied the safety function of 
that control rod since it is in a position of maximum contribution 
to shutdown capability. A plant shutdown for this situation would 
result in an unnecessary plant thermal cycle without any 
compensatory safety benefit. Under the proposed change, inoperable 
inserted rods would continue to be counted in the operability 
requirement precluding power operation with more than 8 inoperable 
control rods.
    The second proposed change removes the control rod drive (CRD) 
water pressure limits from the insertion capability test of 
inoperable, non-stuck, control rods. Reactor pressure, assisted by a 
pre-charged accumulator, provides the driving force for the rapid 
shutdown of the reactor (scram), independent of the CRD water 
pressure. Variation of this pressure is not an indicator of a 
degraded control rod, and does not inhibit the safety function of 
the control rod. Control rod scram and exercise testing requirements 
assure the operability of the CRD system. The proposed change would 
eliminate the need to unnecessarily insert a control rod into the 
core if it could not be repositioned using the normal drive water 
pressure setting.
    The third proposed change would limit the increased frequency 
surveillance requirement (every 24 hours) exercise test of withdrawn 
control rods upon discovery of an immovable control rod to a one-
time test in lieu of every 24 hours. A one-time 24-hour test is 
sufficient to determine if a generic control rod problem exists. 
Under the proposed change, following the 24-hour test, and in 
absence of any additional detectable problems, the control rod 
exercise test would revert back to a normal testing frequency. 
Repetitive 24-hour tests [have] the potential to reduce the operable 
lifespan of hydraulic control unit components and increases the 
potential for a reactivity management event.
    The proposed changes will not impede the ability of the 
surveillance requirements to detect control rod degradation, or 
inhibit the control rod drive system from performing its designed 
safety function.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No. The proposed changes do not alter the physical 
design, safety limits, or safety analysis assumptions, associated 
with the operation of the plant. Accordingly, the changes do not 
introduce any new accident initiators, nor do they reduce or 
adversely affect the capabilities of any plant structure, system, or 
component to perform their safety function.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No. A fully inserted [control] rod has satisfied its 
safety function by being in the position of maximum contribution to 
shutdown reactivity. Eliminating the CRD water pressure limits does 
not impact scram capability. Further, the proposed changes will 
eliminate extended accelerated control rod testing that may shorten 
the lifespan of control components without any compromise in the 
detection of control rod operability problems. The proposed changes 
would not impact control rod operability and surveillance 
requirements that are necessary to assure that the control rod 
system will perform its designed safety function.
    Therefore, the proposed amendment 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. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.

[[Page 29795]]

    NRC Section Chief: Darrell J. Roberts.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: April 20, 2005.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to replace plant-specific 
position titles with generic position titles. The proposed changes are 
consistent with NUREG-1430, ``Standard Technical Specifications--
Babcock and Wilcox Plants,'' Revision 3. Also, the licensee proposes to 
delete TS 6.7, ``Safety Limit Violation or Protective Limit 
Violation,'' including a change to TS 2.1.2, ``Safety Limits and 
Limiting Safety System Settings--Reactor Core,'' associated with the 
deletion of TS 6.7. Additionally, the licensee proposes to relocate to 
the Technical Requirements Manual (TRM), the Process Control Program 
requirements from TS 6.8, ``Procedures and Programs,'' and from TS 
6.14, ``Process Control Program (PCP).'' Associated with this change, 
TS Definition 1.30, ``Process Control Program,'' is proposed to be 
deleted. Also, TS 6.15, ``Offsite Dose Calculation Manual (ODCM),'' is 
proposed to be modified to eliminate the requirement that changes to 
the ODCM be reviewed and accepted by the Plant Operations Review 
Committee (PORC). Lastly, the licensee proposes to revise in the TS the 
title, ``Industrial Security Plan'' to ``Physical Security Plan.''
    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 changes affect the requirements for the 
administrative controls section of the Technical Specifications. The 
proposed changes are primarily intended to make the plant-specific 
position/organizational titles found in the administrative controls 
section of the Technical Specifications more generic. The proposed 
changes do not affect any plant structures, systems, and components, 
and have no effect on plant operations. The proposed changes are 
administrative and do not affect any existing limits. Accident 
initial conditions, probability, and assumptions remain as 
previously analyzed. The proposed changes will have no effect on 
accident initiation frequency. The proposed changes do not 
invalidate the assumptions used in evaluating the radiological 
consequences of any accident. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative and do not introduce any 
new or different accident initiators. Therefore, the proposed 
changes do 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 changes are administrative and will not have a 
significant effect on any margin of safety. 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: April 22, 2005.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) related to fuel handling and 
storage. Specifically, the proposed change is to reflect that spent 
fuel storage racks are no longer installed in the cask pit or transfer 
pit and that there are no longer any low-density fuel storage racks in 
the spent fuel pool. Additionally, the proposed changes would relocate 
the requirements of TS 3/4.9.7, ``Crane Travel--Fuel Handling 
Building,'' to the Technical Requirements Manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would relocate the requirements of TS 3/
4.9.7 to the DBNPS [Davis-Besse Nuclear Power Station] TRM. Any 
subsequent changes to the TRM would require evaluation under the 
appropriate regulatory processes (e.g., 10 CFR 50.59). The proposed 
relocation of TS 3/4.9.7 does not affect any accident initiators. 
The relocated TRM requirements will assure the initial conditions 
assumed in the analysis of a fuel handling accident are maintained. 
The proposed change does not affect the ability of plant equipment 
to mitigate the consequences of any accident. The proposed changes 
to reflect that fuel storage racks are no longer installed in the 
cask pit or transfer pit and that low density fuel storage racks are 
no longer installed in the spent fuel pool are consistent with the 
current plant configuration. The proposed changes do not affect any 
accident initiators. The revised requirements will continue to 
assure the capability to mitigate the consequences of a fuel 
handling accident in the fuel storage area. 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 relocation of TS 3/4.9.7 to the TRM does not alter 
the design, operation, or testing of any structure, system, or 
component. The proposed changes to reflect that fuel storage racks 
are no longer installed in the cask pit or transfer pit and that low 
density fuel storage racks are no longer installed in the spent fuel 
pool are consistent with the current plant configuration. No new 
accident initiators are created. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed relocation of TS 3/4.9.7 to the TRM does not alter 
the design, operation, or testing of any structure, system, or 
component. The proposed changes to reflect that fuel storage racks 
are no longer installed in the cask pit or transfer pit and that low 
density fuel storage racks are no longer installed in the spent fuel 
pool are consistent with the current plant configuration and do not 
adversely affect the ability of any structure, system, or component 
to perform its safety function. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.

[[Page 29796]]

    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: May 2, 2005.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) Figure 2.1-1, ``Reactor Core Safety 
Limit'' and TS Table 2.2-1, ``Reactor Protection System Instrumentation 
Trip Setpoints.'' These TS revisions would support the use of Framatome 
Mark B-HTP fuel in the reactor.
    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 changes include a revision of the Reactor Core 
Safety Limits specified in Technical Specification (TS) Section 
2.1.1, and a revision of the Reactor Protection System (RPS) Reactor 
Coolant System (RCS) Pressure-Temperature setpoint Allowable Value 
provided in TS Section 2.2.1. The proposed changes preserve the 
design DNB [departure from nucleate boiling] Ratio safety criterion 
that there shall be at least a 95% probability at a 95% confidence 
level that the hot fuel rod in the core does not experience a 
departure from nucleate boiling during normal operation or events of 
moderate frequency. Further, there are no evaluated accidents in 
which the fuel cladding or fuel assembly structural components are 
assumed to arbitrarily fail as an accident initiator. The fuel 
handling accident analysis assumes that the cladding does, in fact, 
fail as a result of an undefined fuel handling event. However, the 
probability of an accident initiator for the fuel handling accident 
is independent of the parameters changed in this amendment request. 
In addition, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated 
because the proposed changes do not alter any assumptions previously 
made in the radiological consequence evaluations, or affect 
mitigation of the radiological consequences of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because no new accident scenarios, failure mechanisms or single 
failures are introduced as a result of the proposed. All systems, 
structures, and components previously required for the mitigation of 
an event remain capable of fulfilling their intended design 
function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety because extensive analyses of the primary fission 
product barriers, conducted in support of the proposed changes, have 
concluded that all relevant design criteria remain satisfied, both 
from the standpoint of the integrity of the primary fission product 
barrier and from the standpoint of compliance with the regulatory 
acceptance criteria. As appropriate, all evaluations have been 
performed using methods that have either been reviewed and approved 
by the Nuclear Regulatory Commission or that are in compliance with 
applicable regulatory review guidance and standards.
    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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: January 10, 2005.
    Description of amendment request: The amendment request proposes to 
revise the surveillance interval associated with Technical 
Specification Surveillance Requirement 4.6.1.3b from once every 6 
months to once every 24 months for verification that only one door in 
each containment air lock can be opened at a time.
    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 amendment will neither effect nor change any design 
function, or method of performing or controlling design functions, 
or any analysis that verifies the capability of structures, systems 
and components (SSCs) to perform their designed function(s). The 
proposed amendment will have no adverse effect on plant operation or 
its controlled configuration. As a result, the proposed amendment 
will not change assumptions, or change, degrade or prevent actions 
described or assumed in accidents evaluated and described in the 
Seabrook Station Updated Final Safety Analysis Report (UFSAR). The 
proposed change extends the surveillance interval from 6 months to 
24 months to verify proper functioning of the containment air lock 
interlocks. The proposed change to the Surveillance Requirement 
testing interval does not adversely affect performance of the 
Surveillance Requirement that verifies the functional status of the 
air lock interlock to prevent both air lock doors to be open 
simultaneously. Containment integrity is not affected by the 
proposed amendment. The radiological consequences of an event are 
unchanged, since the functional status of the air lock interlock is 
not adversely affected and the air lock doors' ability to withstand 
the maximum expected post accident containment pressure is not 
adversely affected by the proposed change. Therefore, the proposed 
amendment does not adversely affect nuclear safety or continued safe 
operation of Seabrook Station, or result in an increase in the 
radiological consequences of any accident described in the Seabrook 
Station UFSAR.
    Therefore, it is concluded that the proposed change does not 
involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed amendment will neither effect nor change any design 
function, or method of performing or controlling design functions, 
or any analysis that verifies the capability of structures, systems 
and components (SSCs) to perform their designed function(s). The 
proposed amendment will have no adverse effect on plant operation or 
its controlled configuration. As a result, the proposed amendment 
will not change assumptions, or change, degrade or prevent actions 
described or assumed in accidents evaluated and described in the 
Seabrook Station UFSAR. There are no changes associated with 
extending the surveillance interval for the air lock interlock that 
could potentially introduce new failure modes or accident 
initiators.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change extends the surveillance interval from 6 
months to 24

[[Page 29797]]

months to verify proper functioning of the containment air lock 
interlock. The containment air lock interlocks are normally not 
challenged and operating experience has shown these components have 
an excellent surveillance pass rate. Furthermore, increasing the 
surveillance interval has no affect on the air lock doors' ability 
to withstand the maximum expected post accident containment 
pressure. Containment integrity is not affected by the proposed 
amendment. The proposed amendment will neither effect nor change any 
design function, or method of performing or controlling design 
functions, or any analysis that verifies the capability of 
structures, systems and components (SSCs) to perform their designed 
function(s). The functional status of the containment air lock 
interlocks will continue to be verified.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in the margin of safety.

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

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 28, 2005.
    Description of amendment request: The proposed amendment would 
extend the expiration date of Facility Operating License (FOL) NPF-86 
for Seabrook Station, Unit No. 1 by approximately 3.4 years. The 
extension would set the date of expiration of the FOL to occur 40 years 
from the date of issuance of the full-power operating license. 
Specifically, the FOL, with a current expiration date of October 17, 
2026 would be revised to expire on March 15, 2030. This change would 
allow the recapture of zero-power and low-power testing time in 
accordance with SECY-98-296, ``Agency Policy Regarding Licensee 
Recapture of Low-Power Testing or Shutdown Time for Nuclear Power 
Plants,'' dated December 21, 1998.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated since it does not involve a change to design configuration 
or operation of the facility. The proposed change does not effect 
the source term, containment isolation or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated in the Seabrook Station UFSAR [updated 
final safety analysis report]. In addition, Seabrook Station Unit 
[No.] 1 was designed and constructed to ensure a 40-year service 
life. Design features were incorporated that provide for inspection 
of structures, systems and components during the 40-year service 
life. Surveillance, inspection and maintenance practices have been 
implemented in accordance with the American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code and the unit Technical 
Specifications to provide assurance that any degradation in plant 
safety-related equipment will be identified and corrected to provide 
continued safe operation of the unit throughout the duration of the 
facility operating license.
    The recapture period requested by this amendment is for 3.4 
years. This time is insignificant from an aging effect perspective 
when considered in conjunction with the surveillance, inspection and 
maintenance programs implemented to provide early indication of 
degradation in plant safety-related equipment. Continual maintenance 
and testing provides for continued safe operation of the unit 
throughout the duration of the facility operating license.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed amendment revises the expiration of the facility 
operating license such that the expiration of the facility operating 
license is based upon issuance of the FPOL [full-power operating 
license] and not upon issuance of the ZPOL/LPOL [zero-power 
operating license/low-power operating license]. The proposed 
change[s] do[es] not involve physical alteration of plant systems[,] 
structures or components or changes in parameters governing the 
manner in which the plant is operated and maintained.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed amendment revises the expiration of the facility 
operating license such that the expiration of the facility operating 
license is based upon issuance of the FPOL and not upon issuance of 
the ZPOL/LPOL. No physical changes are being made to the design 
features or operation of the facility.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary and the containment structure) to limit the 
radiological dose to the public and control room operators in the 
event of an accident. The proposed amendment to the facility 
operating license has no impact on the margin of safety and 
robustness provided in the design and construction of the facility. 
In addition, the proposed amendment will not relax any of the 
criteria used to establish safety limits, nor will the proposed 
amendment relax safety system settings or limiting conditions of 
operation as defined in the Technical Specifications.
    Therefore, the proposed amendment does not result in a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: April 26, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 5.6.5.b., ``Core Operating Limits 
Report (COLR),'' to add the Palisades-specific fuel assembly growth 
model to the analytical methods referenced in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment augments an existing analytical 
method used to determine the core operating limits per Technical 
Specification 5.6.5.b. Accidents previously evaluated will be 
unaffected because they will continue to be analyzed using 
applicable methodologies approved by the Nuclear Regulatory 
Commission to ensure all required safety limits are met. The 
proposed amendment does not affect the acceptance criteria for any 
Final Safety Analysis Report (FSAR) safety analysis analyzed 
accidents and anticipated operational occurrences. As such, the 
proposed amendment does not increase the probability or consequences 
of an accident. The proposed amendment does not involve

[[Page 29798]]

operation of the required structures, systems or components (SSCs) 
in a manner or configuration different from those previously 
recognized or evaluated.
    Therefore, 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.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
any SSC or a change in the way any SSC is operated. The proposed 
amendment does not involve operation of any required SSCs in a 
manner or configuration different from those previously recognized 
or evaluated. No new failure mechanisms will be introduced by the 
changes being requested.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not, by itself, introduce a failure 
mechanism. The proposed amendment does not involve any physical 
changes to the plant or manner in which the plant is operated. The 
proposed changes do not affect the acceptance criteria for any FSAR 
safety analysis analyzed accidents or anticipated operational 
occurrences. All required safety limits would continue to be 
analyzed using methodologies approved by the Nuclear Regulatory 
Commission.
    Therefore, the proposed amendment would 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: November 23, 2004.
    Description of amendment request: The proposed amendment revises 
the descriptive wording of Technical Specifications Table 1-1, ``RPS 
[reactor protection system] Limiting Safety System Settings,'' for the 
Reactor Trip setpoint for Low Steam Generator Water Level to relocate 
unnecessary detail and converts Technical Specifications Section 4.0, 
Design Features, to be consistent with NUREG-1432, Revision 3, 
``Standard Technical Specifications for Combustion Engineering 
Plants.'' These changes will be needed to support the operation of Fort 
Calhoun Station (FCS) after major components (steam generators, 
pressurizer, and reactor vessel head) are replaced in 2006.
    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 changes are not related to an initiator of any 
previously evaluated accident. The proposed changes revise 
descriptive information only, and will not prevent safety systems 
from performing their accident mitigation function as assumed in the 
safety analysis.
    Therefore, this change does not involve a significant increase 
in the probability or 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 changes only relocate descriptive information in 
the Technical Specifications to the USAR [Updated Safety Analysis 
Report]. Modifications will not be made to existing equipment nor 
will any new or different types of equipment be installed. The 
proposed changes to the Technical Specifications will not alter 
assumptions made in safety analysis and licensing bases.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed administrative changes only relocate descriptive 
information in the FCS Technical Specifications to the USAR, and 
have no effect on safety margins.
    Therefore, this technical specification change does not involve 
a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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: September 8, 2004.
    Description of amendment request: The proposed amendments would 
change the SSES 1 and 2 Technical Specifications (TSs) limiting 
conditions for operation (LCO) 3.8.4, ``DC Sources-Operating,'' to 
incorporate the Technical Specifications Change Task Force (TSTF) 16, 
Revision 2, and other unrelated editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence [sic] or consequences of an accident 
previously evaluated?
    Response: No.
    The Technical Specification allowed Completion Time for any 
inoperability is not an initiator to any accident sequence analyzed 
in the Final Safety Analysis Report (FSAR). The changes do not 
involve any physical change to structures, systems, or components 
(SSCs) and does not alter the method of operation or control of 
SSCs. The current assumptions in the safety analysis regarding 
accident initiators and mitigation of accidents are unaffected by 
these changes. No additional failure modes or mechanisms are being 
introduced and the likelihood of previously analyzed failures 
remains unchanged.
    Operation in accordance with the proposed Technical 
Specification (TS) ensures that the AC distribution system and 
supported equipment functions remain capable of performing the 
function as described in the FSAR. Therefore, the mitigative 
functions supported by the system will continue to provide the 
protection assumed by the analysis.
    The correction of typographical errors, changes in format and 
the deletion of a no longer required one-time exemption are 
administrative changes.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints, at which protective or mitigative actions are 
initiated, affected by this change. This change will not alter the 
manner in which equipment operation is initiated, nor will the

[[Page 29799]]

function demands on credited equipment be changed. No alterations in 
the procedures that ensure the plant remains within analyzed limits 
are being proposed, and no changes are being made to the procedures 
relied upon to respond to an off-normal event as described in the 
FSAR. The correction of typographical errors, changes in format and 
the deletion of a no longer required one-time exemption are 
administrative changes. As such, no new failure modes are being 
introduced. The change does not alter assumptions made in the safety 
analysis and licensing basis.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change is acceptable because the 
restoration times for deenergized AC distribution subsystems has 
been previously evaluated in Unit 2 Amendment No. 148. Additional 
margin of safety is gained with the inclusion of the requirement to 
enter applicable actions for inoperable Class lE battery chargers as 
a result of inoperable AC bus(es). The correction of typographical 
errors, changes in format and the deletion of a no longer required 
one-time exemption are administrative changes. Therefore the plant 
response to analyzed events will continue to provide the margin of 
safety assumed by the analysis.

    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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: January 28, 2005.
    Description of amendment request: The proposed amendments would 
change the SSES 1 and 2 Technical Specifications (TSs) 5.5.6, 
``Inservice Testing Program,'' to replace the reference to American 
Society of Mechanical Engineers (ASME) Boiler and PressureVessel Code, 
Section XI, with a reference to ASME Code for Operation and Maintenance 
of Nuclear Power Plants (ASME OM Code) as the source of requirements 
for the inservice testing of ASME Code Class 1, 2, and 3 pumps and 
valves. These changes are consistent with the implementation of the 
SSES 1 and 2 Third 10-Year Interval Inservice Testing Program in 
accordance with the requirements of 10 CFR 50.55a(f).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence [sic] or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes revise Technical Specification 5.5.6 for 
SSES Units 1 and 2 to conform to the requirements of 10 CFR 
50.55a(f) regarding the inservice testing of pumps and valves for 
the Third 10-Year Interval. The current Technical Specifications 
reference the ASME Boiler and Pressure Vessel Code, Section XI, 
requirements for the inservice testing of ASME Code Class 1, 2, and 
3 pumps and valves. The proposed changes would reference the ASME OM 
Code, which is consistent with 10 CFR 50.55a(f) and accepted for use 
by the NRC. The proposed changes are administrative in nature.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise Technical Specification 5.5.6 for 
SSES Units I and 2 to conform to the requirements of 10 CFR 
50.55a(f) regarding the inservice testing of pumps and valves for 
the Third 10-Year Interval. The current Technical Specifications 
reference the ASME Boiler and Pressure Vessel Code, Section XI, 
requirements for the inservice testing of ASME Code Class 1, 2, and 
3 pumps and valves. The proposed changes would reference the ASME OM 
Code, which is consistent with 10 CFR 50.55a(f)and accepted for use 
by the NRC. The proposed changes are administrative in nature.
    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 changes revise Technical Specification 5.5.6 for 
SSES Units I and 2 to conform to the requirements of 10 CFR 
50.55a(f) regarding the inservice testing of pumps and valves for 
the Third 10-Year Interval. The current Technical Specifications 
reference the ASME Boiler and Pressure Vessel Code, Section XI, 
requirements for the inservice testing of ASME Code Class 1, 2, and 
3 pumps and valves. The proposed changes would reference the ASME OM 
Code, which is consistent with 10 CFR 50.55a(f) and accepted for use 
by the NRC. The proposed changes are administrative in nature.
    Therefore, the proposed change[s] does [sic] 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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: February 7, 2005.
    Description of amendment request: The proposed amendments would 
change the SSES 1 and 2 Technical Specifications (TSs) for ``Secondary 
Containment,'' limiting condition for operation (LCO) 3.6.4.1, by 
revising the frequency note applicable to Surveillance Requirements 
(SR) 3.6.4.1.4 and SR 3.6.4.1.5. The revised note requires each SR be 
performed with the 3 zone configuration every 60 months.
    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 does not involve a significant increase in 
the probability of an accident previously evaluated because neither 
Secondary Containment nor the Standby Gas Treatment System is an 
initiator of an accident. Both mitigate accident consequences.
    The consequences of a Design Basis Analysis-Loss of Coolant 
Accident (DBA-LOCA) have been evaluated in the FSAR [final safety 
analysis report]. Revising the surveillance frequency to require the 
most limiting configurations to be tested with the 60-month period 
rather than just the three zone configuration provides assurance 
that the most limiting secondary containment configuration is tested 
every 60 months in accordance with the original intent of the 
surveillance frequency. The proposed change also provides added 
assurance of acceptable performance within the analysis assumptions 
of the FSAR. The radiological evaluation of DBA-LOCA doses, 
including doses offsite, control room habitability, and exposures 
for personnel are not impacted.

[[Page 29800]]

    Therefore, this 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 changes do not involve a physical alteration of the 
plant. No new or different [kind] of equipment will be installed nor 
will there be changes in methods governing normal plant operation.
    The potential for the loss of plant systems or equipment to 
mitigate the effects of an accident is not altered.
    The proposed changes do not require any new operator response or 
introduce any new opportunities for operator error not previously 
considered.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in 
[a] margin of safety?
    Response: No.
    The proposed change does not involve a significant reduction in 
[a] margin of safety.
    The surveillance test change ensures all the secondary 
containment configurations are tested within a 60-month period when 
only one configuration was previously required to be tested. This 
change has a positive effect on the margin of safety as it provides 
more restrictive testing requirement that will provide added 
assurance of acceptable secondary containment performance.

    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania

    Date of amendment request: March 18, 2005.
    Description of amendment request: The proposed amendment would 
revise the SSES 2 Technical Specification (TS) 3.3.8.1, ``Loss of Power 
(LOP) Instrumentation,'' to: (1) clarify that Condition A applies to 
inoperable instrumentation other than during the performance of 
Surveillance Requirement (SR) 3.8.1.19 loss-of-coolant accident/loss of 
offsite power testing on Unit 1 and to revise TS Bases section to 
clarify that this condition is applicable to both Unit 1 and Unit 2 LOP 
Instrumentation, (2) add new Condition B to allow the LOP 
instrumentation for two Unit 1 4.16kV Engineered Safeguards System 
buses in the same Division to be inoperable for up to 8 hours for the 
performance of SR 3.8.1.19 on Unit 1. In addition, the proposed 
amendment would revise the SSES 2 TS 3.8.7, ``Distribution Systems-
Operating,'' to: (1) eliminate ``or more'' and the plural to subsystems 
such that the condition would read ``One Unit 1 AC [alternating 
current] electrical power distribution subsystem inoperable,'' (2) add 
new Condition D for two Unit 1 AC electrical power distribution 
subsystems inoperable.
    This will impose an 8-hour Completion Time for restoration of at 
least one of the two Unit 1 AC distribution subsystems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The Technical Specification allowed Completion Time for any 
inoperability is not an initiator to any accident sequence analyzed 
in the Final Safety Analysis Report (FSAR). The changes do not 
involve any physical change to structures, systems, or components 
(SSCs) and does not alter the method of operation or control of 
SSCs. The current assumptions in the safety analysis regarding 
accident initiators and mitigation of accidents are unaffected by 
these changes. No additional failure modes or mechanisms are being 
introduced and the likelihood of previously analyzed failures 
remains unchanged.
    Operation in accordance with the proposed Technical 
Specification (TS) ensures that the AC distribution system and 
supported equipment functions remain capable of performing the 
function as described in the FSAR [final safety analysis report]. 
Therefore, the mitigative functions supported by the system will 
continue to provide the protection assumed by the analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints, at which protective or mitigative actions are 
initiated, affected by this change. This change will not alter the 
manner in which equipment operation is initiated, nor will the 
function demands on credited equipment be changed. No alterations in 
the procedures that ensure the plant remains within analyzed limits 
are being proposed, and no changes are being made to the procedures 
relied upon to respond to an off-normal event as described in the 
FSAR. As such, no new failure modes are being introduced. The change 
does not alter assumptions made in the safety analysis and licensing 
basis.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change is acceptable because the 
restoration time for deenergized AC distribution subsystems has been 
previously evaluated in Unit 2 Amendment No. 148. Therefore[,] the 
plant response to analyzed events will continue to provide the 
margin of safety assumed by the analysis.

    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: March 4, 2005.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3.5.1, ``Accumulators,'' to extend 
the completion time (CT) for Action (a) from 1 hour to 24 hours. The 
accumulators are part of the emergency core cooling system and consist 
of tanks partially filled with borated water and pressurized with 
nitrogen gas. The contents of the tank are discharged to the reactor 
coolant system (RCS) if, as during a loss-of-coolant accident (LOCA), 
the coolant pressure decreases to below the accumulator pressure. 
Action (a) of TS 3.5.1 specifies a CT to restore an accumulator to 
operable status when it has been declared inoperable for a reason other 
than the boron concentration of the water in the accumulator not being 
within the required range. This change was proposed by the Westinghouse 
Owners Group participants in the TS Task Force (TSTF) and is designated 
TSTF-370. TSTF-370 is supported by NRC-

[[Page 29801]]

approved Topical Report WCAP-15049-A, ``Risk-Informed Evaluation of an 
Extension to Accumulator Completion Times,'' submitted on May 18, 1999. 
The NRC staff issued a Notice of Opportunity for Comment in the Federal 
Register on July 15, 2002 (67 FR 46542), on possible amendments 
concerning TSTF-370, 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 March 12, 2003 (68 FR 
11880). The licensee affirmed the applicability of the following NSHC 
determination in its application dated March 4, 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

    The basis for the accumulator limiting condition for operation 
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a 
sufficient volume of borated water will be immediately forced into 
the core through each of the cold legs in the event the RCS pressure 
falls below the pressure of the accumulators, thereby providing the 
initial cooling mechanism during large RCS pipe ruptures. As 
described in Section 9.2 of the WCAP-15049, ``Risk-Informed 
Evaluation of an Extension to Accumulator Completion Times,'' 
evaluation, the proposed change will allow plant operation with an 
inoperable accumulator for up to 24 hours, instead of 1 hour, before 
being required to begin shutdown. The impact of the increase in the 
accumulator CT on core damage frequency for all the cases evaluated 
in WCAP-15049 is within the acceptance limit of 1.0E-06/yr for a 
total plant core damage frequency less than 1.0E-03/yr. The 
incremental conditional core damage probabilities calculated in 
WCAP-15049 for the accumulator CT increase meet the criterion of 5E-
07 in Regulatory Guides (RGs) 1.174 [``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis''] and 1.177 [``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications''] for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049, design 
basis accumulator success criteria are not considered necessary to 
mitigate large-break LOCA events, and were only included in the 
WCAP-15049 evaluation as a worst-case data point. In addition, WCAP-
15049 states that the NRC has indicated that an incremental 
conditional core damage frequency greater than 5E-07 does not 
necessarily mean the change is unacceptable.
    The proposed TS change does not involve any hardware changes nor 
does it affect the probability of any event initiators. There will 
be no change to normal plant operating parameters, engineered safety 
feature actuation setpoints, accident mitigation capabilities, 
accident analysis assumptions or inputs.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2: The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this 
proposed TS CT increase. All safety systems still function in the 
same manner and there is no additional reliance on additional 
systems or procedures. The proposed accumulator CT increase has a 
very small impact on core damage frequency. The WCAP-15049 
evaluation demonstrates that the small increase in risk due to 
increasing the accumulator allowed outage time is within the 
acceptance criteria provided in RGs 1.174 and 1.177. No new 
accidents or transients can be introduced with the requested change 
and the likelihood of an accident or transient is not impacted.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be caused as a result 
of the proposed TS change. No new failure mode has been created and 
no new equipment performance burdens are imposed.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

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

    The proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits.
    The basis for the accumulator LCO, as discussed in Bases Section 
3.5.1, is to ensure that a sufficient volume of borated water will 
be immediately forced into the core through each of the cold legs in 
the event the RCS pressure falls below the pressure of the 
accumulators, thereby providing the initial cooling mechanism during 
large RCS pipe ruptures. As described in Section 9.2 of the WCAP-
15049 evaluation, the proposed change will allow plant operation 
with an inoperable accumulator for up to 24 hours, instead of 1 
hour, before being required to begin shutdown. The impact of this on 
plant risk was evaluated and found to be very small. That is, 
increasing the time the accumulators will be unavailable to respond 
to a large LOCA event, assuming accumulators are needed to mitigate 
the design basis event, has a very small impact on plant risk. Since 
the frequency of a design basis large LOCA (a large LOCA with loss 
of offsite power) would be significantly lower than the large LOCA 
frequency of the WCAP-15049 evaluation, the impact of increasing the 
accumulator CT from 1 hour to 24 hours on plant risk due to a design 
basis large LOCA would be significantly less than the plant risk 
increase presented in the WCAP-15049 evaluation.

    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the NRC staff proposes to 
determine that the requested change does not involve a 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.

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

    Date of amendment request: August 5, 2004, as superceded in its 
entirety by letter dated March 15, 2005.
    Brief description of amendments: The proposed amendments would 
revise Technical Specification (TS) 3.7.10 entitled ``Control Room 
Emergency Filtration/Pressurization System (CREFS)'' to extend the 
Completion Time for ACTION B., ``Two CREFS Trains inoperable due to 
inoperable Control Room boundary in MODES 1, 2, 3, and 4'' from 24 
hours to 14 days for implementation of the Turbine Generator Protection 
System Digital Modification currently scheduled during the eleventh 
refueling outage for Unit 1 (1RF11) and the ninth refueling outage for 
Unit 2 (2RF09). The description of CONDITION E would also be revised 
for implementation of this modification.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.

    1. Do the proposed changes involve a significant increase the 
probability or consequences of an accident previously evaluated?
    Response: No.

[[Page 29802]]

    This is a revision to the Technical Specifications for the CREFS 
which is a mitigation system designed to minimize in leakage and to 
filter the Control Room atmosphere to protect the operator following 
accidents previously analyzed. An important part of the system is 
the Control Room boundary. The Control Room boundary integrity is 
not an initiator or precursor to any accident previously evaluated. 
Therefore, the probability of any accident previously evaluated is 
not increased. The analysis of the consequences of analyzed accident 
scenarios under the Control Room breach conditions along with the 
compensatory actions for restoration of Control Room integrity 
demonstrate that the consequences of any accident previously 
evaluated are not increased. Therefore, it is concluded that this 
change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not impact the accident analysis. The 
change will not alter the requirements of the CREFS or its function 
during accident conditions. The administrative controls and 
compensatory actions will ensure the CREFS will perform its safety 
function. No new or different accidents result from the revised 
Completion Time or the restated TS Condition E. The change does not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. The change does not alter 
assumptions made in the safety analysis. The proposed change is 
consistent with the safety analysis assumptions and current plant 
operating practice. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by these changes. The proposed change will not 
result in plant operation in a configuration outside the design 
basis for an unacceptable period or time without compensatory 
actions and administrative controls. The proposed change does not 
affect systems that respond to safely shutdown the plant and to 
maintain the plant in a safe shutdown condition. Therefore the 
proposed change does not involve a reduction in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92'') are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Allen G. Howe.

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: September 15, 2004.
    Brief description of amendment: The amendment deleted the Technical 
Specification (TS) requirements related to hydrogen recombiners and 
hydrogen/oxygen monitors. The TS changes are consistent with the 
revision of Title 10, Code of Federal Regulations, Section 50.44, 
``Standards for Combustible Gas Control System in Light-Water-Cooled 
Power Reactors,'' that became effective on October 16, 2003; and 
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: April 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 164.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 01, 2005 (70 
FR 5235). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 28, 2005.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: December 1, 2004.
    Brief description of amendments: The amendments eliminate the 
requirements to submit monthly operating reports and occupational 
radiation exposure reports.
    Date of issuance: May 9, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 272 and 249.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5236).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated May 9, 2005.
    No significant hazards consideration comments received: No.

[[Page 29803]]

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

    Date of application for amendment: December 6, 2004.
    Brief description of amendment: The amendment deleted the 
requirements to submit monthly operating reports and occupational 
radiation exposure reports.
    Date of issuance: April 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 166.
    Facility Operating License No. NPF-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5236).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: July 19, 2004, as supplemented 
by letters dated March 8 and March 22, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 3.8.4, ``DC Sources--Operating'' and TS 
3.8.6, ``Battery Cell Parameters'' to allow for the replacement of the 
existing nickel-cadmium diesel generator batteries with conventional 
lead-acid batteries.
    Date of issuance: April 27, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance April 27, 2005.
    Amendment Nos.: 223 and 218.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 2004 (69 
FR 76488). The supplements dated March 8 and March 22, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the July 19, 2004, 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 April 27, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 9, 2004, and as 
supplemented by letter dated April 1, 2005.
    Brief description of amendment: This amendment revises Technical 
Specifications (TS) Limiting Condition for Operation (LCO) 3.4.11, 
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,'' 
to replace the P/T curves for inservice leak and hydrostatic testing, 
non-nuclear heating and cooldown, and nuclear heating and cooldown 
currently illustrated in TS Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3, 
respectively.
    Date of issuance: May 12, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 193.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53102).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 12, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 17, 2004.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,'' 
and TS 5.6.4, ``Monthly Operating Reports.''
    Date of issuance: May 3, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 167.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9992).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: April 14, 2004, as supplemented 
on December 15, 2004.
    Brief description of amendment: This amendment eliminates secondary 
containment operability requirements when handling sufficiently decayed 
irradiated fuel or performing core alterations. The secondary 
containment is still required to be operable during operations with the 
potential to drain the reactor vessel.
    Date of issuance: April 28, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 215.
    Facility Operating License No. DPR-35: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 12, 2004 (69 FR 
60679). The December 15, 2004, supplement provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated April 28, 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: July 17, 2004, as supplemented by 
letters dated October 18, 2004, February 2, February 21, March 8, and 
April 5, 2005.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.3.1, to allow the use of a limited number of lead 
test assemblies, the use of ZIRLOTM as an acceptable fuel 
cladding, and to allow a limited substitution of zirconium alloy or 
stainless steel filler rods for fuel rods, while relocating the maximum 
fuel enrichment from TS 5.3.1 to TS 5.6.1. TS 6.9.1.11.1 is revised to 
allow the use of the Westinghouse Nuclear Physics code package and to 
incorporate the methodology used to support ZIRLOTM cladding 
material. Additionally, the amendment approved the administrative 
changes of correcting a referencing report error of the CESEC code and 
deleting the TS Index from the TSs.

[[Page 29804]]

    Date of issuance: May 9, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 200.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 2004 (69 FR 
43460). The supplements dated October 18, 2004, February 2, February 
21, March 8, and April 5, 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 amendment 
is contained in a Safety Evaluation dated May 9, 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

Docket Nos. 50-010, 50-237 and 50-249, Dresden Nuclear Power Station, 
Units 1, 2 and 3, Grundy County, Illinois

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 
1 and 2, Rock Island County, Illinois

Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 
2, Lake County, Illinois

    Date of application for amendments: October 21, 2004, as 
supplemented January 4, 2005.
    Description of amendments requests: The amendment deletes the TS 
requirements to submit monthly operating reports and annual 
occupational radiation exposure reports. The change is consistent with 
Revision 1 of NRC-approved Technical Specifications Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-369, 
``Elimination of Requirements for Monthly Operating Reports and 
Occupational Radiation Exposure Reports.'' 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: April 29, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: Byron Station, Unit 1--142, Unit 2--142; Braidwood 
Station, Unit 1--136, Unit 2--136; Dresden Nuclear Power Station, Unit 
1--42, Unit 2--214, Unit 3--206; LaSalle County Station, Unit 1--173, 
Unit 2--159; Quad Cities Nuclear Power Station, Unit 1--225, Unit 2--
220; Zion Nuclear Power Station, Unit 1--184, Unit 2--171.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77, 
DPR-2, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29 and DPR-30: The 
amendments revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Date of initial notice in Federal Register: 
April 08, 2005 (70 FR 18061). The notice provided an opportunity to 
submit comments on the Commission's proposed NSHC determination. No 
comments have been received.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 29, 2005.
    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.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendment: October 21, 2004.
    Brief description of amendment: The amendments deleted the 
Technical Specifications (TSs) 6.9.1.5.a and 6.9.1.6 requirements to 
submit monthly operating reports and annual occupational radiation 
exposure reports. The change is consistent with Revision 1 of the U.S. 
Nuclear Regulatory Commission's Technical Specifications Task Force 
(TSTF) Change Traveler, TSTF-369, ``Elimination of Requirements for 
Monthly Operating Reports and Occupational Radiation Exposure 
Reports.''
    Date of issuance: April 29, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 175 and 137.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the TSs.
    Date of initial notice in Federal Register: June 23, 2004 (69 FR 
35067). This TS improvement was announced in the Federal Register as 
part of the Consolidated Line Item Improvement Process. A notice for 
these TS changes was announced on April 8, 2005 (70 FR 18059). The 
April 8, 2005, notice incorrectly referenced a January 4, 2005, 
supplement to the application. This supplement was reference by error. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 29, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 10, 2004.
    Brief description of amendment: This amendment deletes the 
Technical Specifications associated with hydrogen recombiners and 
hydrogen monitors.
    Date of issuance: April 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 135.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7767). Add the following statement, if appropriate.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: January 28, 2004, as 
supplemented by letter dated November, 22, 2004.
    Brief description of amendment: This amendment revised technical 
specifications (TSs) 1.4, ``Frequency,'' 5.5.2, ``Primary Coolant 
Sources Outside Containment,'' and 5.5.11, ``Safety Function 
Determination Program,'' by adopting three industry-proposed Standard 
Technical Specifications (STS) changes, which the Nuclear Regulatory 
Commission (NRC) has approved and included in Revision 3 of the STSs. 
These changes are Technical Specifications Task Force (TSTF) traveler 
numbers 273, 284, and 299. The licensee's request to revise TS 3.3.1.1, 
``Reactor Protection System Instrumentation,'' which is associated with 
TSTF-264 is addressed by the NRC staff by a separate Safety Evaluation.
    Date of issuance: May 12, 2005.

[[Page 29805]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 258.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19571).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 12, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: February 10, 2004.
    Brief description of amendments: The amendments (1) extended from 1 
hour to 24 hours the completion time (CT) for Condition C of technical 
specification (TS) 3.5.1, which defines requirements for the safety 
injection accumulators. Condition C of TS 3.5.1 specifies a CT to 
restore an accumulator to operable status when it has been declared 
inoperable for a reason other than the boron concentration of the water 
in the accumulator not being within the required range; (2) deleted 
Condition B which permits one or both accumulators to be inoperable, by 
removing power to the accumulator isolation valve(s), for maintenance 
or testing; (3) modified Condition E to remove reference to Condition 
B; and (4) re-lettered the Conditions and Actions to reflect deletion 
of Condition B.
    Date of issuance: April 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 217, 222.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19573).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 28, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 3, 2004, as supplemented by 
letters dated February 4, and March 28, 2005.
    Brief description of amendments: The amendments revise the 
licensing to define a new hydraulic analysis methodology for 
demonstrating functionality of the cooling water (CL) system following 
a design-basis seismic event. The seismic analysis methodology for the 
CL system is revised to include (1) evaluation of CL system performance 
following a seismic event assuming a rupture of a non-seismic pipe at 
the worst case location, and (2) application of acceptance criteria 
from the American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code, Section lll, to demonstrate that the CL system non-seismic 
piping will maintain pressure boundary integrity with design-basis 
seismic loads.
    Date of issuance: May 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 169, 159.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40677).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 10, 2005.
    No significant hazards consideration comments received: No.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: August 6, 2004, as supplemented 
March 14, 2005.
    Brief description of amendment: This amendment deletes the 
Technical Specification requirements associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: May 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 90.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7768). The supplement dated March 14, 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 5, 2005.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: December 27, 2004.
    Brief description of amendments: The amendments delete TS 
5.7.1.1.a, ``Occupational Radiation Exposure Report'' and TS 5.7.1.4, 
``Monthly Operating Reports.''
    Date of issuance: May 10, 2005.
    Effective date: May 10, 2005, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2--195; Unit 3--186.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5248). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 10, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendment: December 2, 2004, as 
supplemented by letters dated February 15, March 9, and April 11, 2005.
    Brief description of amendment: The amendment revises portions of 
the Sequoyah Unit 2 Technical Specification Surveillance Requirement 
4.4.5 to eliminate the requirement to inspect a portion of the tube 
within the tubesheet region. This will allow any flaws in the region, 
which is no longer inspected, to remain in service.
    Date of issuance: May 3, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 291.
    Facility Operating License No. DPR-79: Amendment revises the 
technical specifications.

[[Page 29806]]

    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2899). The supplemental letters provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 28, 2004.
    Brief description of amendments: This amendment deletes the 
Technical Specifications associated with hydrogen recombiners and 
hydrogen monitors.
    Date of issuance: April 21, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 117/117.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7770).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 21, 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 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

[[Page 29807]]

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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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).

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: April 18, 2005, as supplemented by 
letter dated April 19, 2005.
    Description of amendment request: The amendment revises Technical 
Specification (TS) 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' to add changes to the SG inspection scope for Wolf Creek 
Generating Station for only the current refueling outage 14 and the 
subsequent operating cycle. Specifically, the amendment modifies the 
inspection requirements for portions of the SG tubes within the hot leg 
tubesheet region of the SGs.
    Date of issuance: April 28, 2005.
    Effective date: Effective the date of issuance, and shall be 
implemented before entry into Mode 4 in the restart from the current 
Refueling Outage 14.
    Amendment No.: 162.
    Facility Operating License No. NPF-42: Amendment revises the 
technical specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. The Coffey County Republican on April 22 and 
26, 2005, and the Emporia Gazette on April 25 and 26, 2005. The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. Comments have been received. The resolution of the 
comments, the Commission's related evaluation of the amendment, finding 
of exigent circumstances, state consultation, and final NSHC

[[Page 29808]]

determination are contained in a safety evaluation dated April 28, 
2005.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

    Dated at Rockville, Maryland, this 16th day of May, 2005.
    For the Nuclear Regulatory Commission.
James E. Lyons,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 05-10063 Filed 5-23-05; 8:45 am]
BILLING CODE 7590-01-P