[Federal Register Volume 69, Number 234 (Tuesday, December 7, 2004)]
[Notices]
[Pages 70712-70727]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-26606]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly

[[Page 70713]]

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 November 5, 2004, through November 24, 2004. 
The last biweekly notice was published on November 23, 2004 (69 FRN 
68180).

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. (Note: 
Public access to ADAMS has been temporarily suspended so that security 
reviews of publicly available documents may be performed and 
potentially sensitive information removed. Please check the NRC Web 
site for updates on the resumption of ADAMS access.) 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/. (Note: Public access to ADAMS has been 
temporarily suspended so that security reviews of publicly available 
documents may be performed and potentially sensitive information 
removed. Please check the NRC Web site for updates on the resumption of 
ADAMS access.) If a request for a hearing or petition for leave to 
intervene is filed within 60 days, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the

[[Page 70714]]

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 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. (Note: Public access to ADAMS has been 
temporarily suspended so that security reviews of publicly available 
documents may be performed and potentially sensitive information 
removed. Please check the NRC Web site for updates on the resumption of 
ADAMS access.) If you do not have access to ADAMS or if there are 
problems in accessing the documents located in ADAMS, contact the NRC 
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: October 20, 2004.
    Description of amendment request: The proposed amendment would 
revise the containment hatch closure requirement in the Technical 
Specifications (TSs) during fuel handling and refueling operations. 
Specifically, the requirement of TS 3.8.6 that the containment 
equipment hatch remain closed with a minimum of 4 bolts securing it in 
place is replaced with the requirement that the equipment hatch be 
capable of being closed during fuel handling and refueling operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change is only related to a postulated fuel 
handling accident inside the Reactor Building occurring during fuel 
loading and refueling activities. The proposed change does not 
increase the probability of a fuel handling accident in that the 
proposed change deals with the results of such an accident, not the 
cause of such an accident. The proposed change does not increase the 
consequences of an accident previously evaluated in that the TMI 
Unit 1 Alternative Source Term has been previously reviewed and 
approved by the NRC [Nuclear Regulatory Commission], and this 
proposed change is consistent with the assumptions of [that] 
previous analysis. The Alternative Source Term analysis for the Fuel 
Handling Accident [i]nside the Reactor Building takes no credit for 
the closure of the containment equipment hatch opening or for a 
filtered release. Previous analyses of external events were reviewed 
and the proposed [change does] not affect the conclusions of these 
analyses. 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 proposed change does not affect nor create a different 
[kind] of fuel handling accident. The proposed change is consistent 
with the existing licensing basis accident analysis for a postulated 
fuel handling accident inside containment during fuel loading and 
refueling operations. The proposed change does not involve any 
structure, system, or component relied upon to mitigate any design 
basis accident. The revised operations are consistent with the fuel 
handling accident analysis. 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Previously approved analysis demonstrates that the resultant 
dose consequences are well within the appropriate acceptance 
criteria. The proposed change is bounded by the previously approved 
analysis, and thus the margin of safety, as defined by 10 CFR 50.67 
and Regulatory Guide 1.183, is maintained. Maintaining the 
capability to close the containment equipment hatch opening 
following an evacuation of the containment would further reduce the 
dose consequences in the event of a fuel handling accident inside 
containment and provides additional margin to the calculated doses. 
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

[[Page 70715]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville, 
IL 60555.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: July 13, 2004.
    Description of amendment request: The proposed amendment would 
revise the fire protection license condition consistent with the 
guidance provided in Generic Letter 88-12, ``Removal of Fire Protection 
Requirements from 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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed license amendment is an administrative change. The 
proposed change will revise the fire protection license condition 
consistent with the guidance provided in Generic Letter 88-12. This 
revision to the fire protection license condition was to be made at 
the time the fire protection requirements were relocated from the 
Technical Specifications to licensee controlled documents. However, 
this change was not requested, nor granted in License Amendment 
Request dated December 4, 1996, approved in Amendment Nos. 227 and 
201. Therefore, the necessary change was not reflected in the 
Operating Licenses.
    This administrative request does not impact the probability or 
consequences of an accident previously evaluated. The incorporation 
of the requested change requires that an evaluation be performed to 
determine the need for prior NRC approval for changes to the Fire 
Protection Program. Changes to administrative programs will result 
from the addition of this condition in the Operating License. 
However, no changes to the facility or the way it is operated are 
expected to result from this change.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed change is administrative. This request does not 
involve a change in the operation of the plant, and no new accident 
initiation mechanism is created by the proposed change, nor does the 
change involve a physical alteration of the plant.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The fire protection requirements were removed from the Technical 
Specifications in accordance with Generic Letter 88-12 in Amendment 
Nos. 227 and 201, with the exception of the change to the Operating 
License's fire protection license condition. The proposed 
administrative change will require an evaluation be performed to 
determine the need for prior NRC approval for changes to the Fire 
Protection Program. No margin of safety is impacted by the proposed 
administrative change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: September 7, 2004.
    Description of amendment request: The proposed changes would allow 
performance of testing for nozzle containment spray blockage to be 
based on the occurrence of activities that could cause nozzle blockage 
rather than a fixed periodic basis. Currently, the testing for nozzle 
blockage is performed every 10 years and Dominion Nuclear Connecticut, 
Inc. (DNC) proposes to change this frequency to ``following maintenance 
that could cause nozzle blockage''. In addition, specific details 
limiting the testing method to an air or smoke test that are currently 
part of the surveillance requirements would be removed. The Technical 
Specification Bases section would be updated with applicable spray 
nozzle testing information and will be expanded to include visual 
inspection.
    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 spray nozzles and the associated containment spray systems 
are designed to perform accident mitigation functions only. The QSS 
[quench spray system] and RSS [recirculation spray system] and 
associated components are not considered as initiators of any 
analyzed accidents. The proposed change does not modify any plant 
equipment and only changes the frequency for performance of a 
surveillance test which does not impact any failure modes that could 
lead to an accident. Removing the testing details from the 
surveillance does not change the ability of the spray nozzles to 
function as assumed and therefore there is no affect on the 
consequence of any accident. Also the proposed change does not 
impact the capability of the QSS and RSS to perform accident 
mitigation functions and therefore does not impact the consequences 
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 QSS and RSS are not being physically modified and there is 
no impact on the capability of the systems to perform accident 
mitigation functions. No system setpoints are being modified and no 
changes are being made to the method in which borated water is 
delivered to the spray nozzles. The testing requirements imposed by 
this proposed change to check for nozzle blockage following 
activities that could cause nozzle blockage do not introduce new 
failure modes for the system. By removing the testing details from 
the surveillance requirement, additional flexibility in the testing 
methodology is allowed for verifying the nozzles are unobstructed 
and assists in ensuring operability of the systems. 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 proposed change does not change or introduce any new 
setpoints at which mitigating functions are initiated. No changes to 
the design parameters of the QSS and RSS are being proposed. No 
changes in system operation are being proposed by this change that 
would impact an established safety margin. The proposed change 
modifies the frequency for verification of nozzle operability in 
such a way that continued high confidence exists for the containment 
spray systems to functions as designed. In addition, removing 
specific testing details from the surveillance does not affect the 
ability of the

[[Page 70716]]

spray nozzles to function as designed. Therefore, based on the 
above, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    In summary, DNC concludes that the proposed amendment does not 
represent a significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c).

    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: Daniel S. Collins, Acting Section Chief.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: November 1, 2004.
    Description of amendment request: The proposed amendment would 
allow the use of a new gantry crane as part of the cask handling system 
in the fuel storage building (FSB) for moving spent fuel casks up to 
110 tons into and out of the spent fuel pit.
    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 amendment would allow the use of the new [IP2] 
FSB gantry crane for loads up to 110 tons, and the new crane will 
prevent the load from being dropped given a single malfunction or 
failure of a portion of the crane. The handling of a loaded spent 
fuel cask is below the maximum load that the crane is designed to 
handle.
    This change does not increase the probability of an accident 
previously evaluated because the probability of a load drop is 
eliminated. The new crane system is designed in accordance with 
NUREG-0554 and Ederer's Generic Licensing Topical Report EDR-1 (NP)-
A, that if a portion of the crane lifting devices malfunctions or 
fails, the load will move a limited distance downward prior to 
backup restraints becoming engaged. The change does not increase the 
consequences of an accident.
    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. The process for transporting a cask with the new crane is 
limited from the FSB truck bay floor to the cask pit area of the 
spent fuel pool. Once a cask is loaded with spent fuel, it is lifted 
from the spent fuel pit, and lowered into the truck bay. The cask is 
never carried over spent fuel in the spent fuel pit.
    Therefore, the change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The [IP2] FSB gantry crane has been installed to comply with 
the single-failure-proof requirements of NUREG-0554 and NRC-approved 
Ederer Topical Report EDR-1, Revision 3, dated October 8, 1982. The 
installation provides additional load carrying capability up to 110 
tons and additional safety features to prevent a cask drop. The 
safety margins provided by the new crane prevent failure of the 
crane or any lifting devices associated with it. The implementation 
of NUREG-0612 general guidelines for the FSB gantry crane provides 
further assurance that safe load paths, procedures, crane operator 
training, and crane inspection and maintenance activities will be 
established to ensure crane operation is performed in a consistently 
reliable manner.
    Therefore, operation of the facility in accordance with 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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 29, 2004.
    Description of amendment request: The proposed change to the 
Technical Specification (TS) assures that sufficient fuel oil 
inventories are available in the Emergency Diesel Generator (DG) Fuel 
Oil Storage Tank (FOST) to support the Extended Power Uprate (EPU) 
consistent with the current licensing basis.
    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, which accounts for the fuel oil consumption 
related to the EPU, will revise the minimum TS volumes associated 
with the DG FOST. The change continues to assure that each DG can 
provide on-site power in the event of an accident and thereby assist 
in the mitigation of the accident.
    The proposed change to the five day full load fuel oil volume 
results in a usable volume 37,000 gallons of fuel oil. The proposed 
change removes the unusable volume (760 gallons) and other 
conservatism (240 gallons) that were included in the current TS. The 
fuel oil volume continues to allow for a runtime of 5 days at full 
load with the removal of this conservatism.
    These changes will not affect the capability of the AC 
[alternate current] Sources to power the systems required to safely 
shutdown the plant. The proposed changes are not accident initiators 
nor do they adversely affect accident initiators or precursors. 
These changes do not affect the mitigation of any accident nor do 
they adversely affect structures, systems, or components that are 
utilized for the mitigation of any analyzed events. The proposed 
changes will have no affect on the radiological consequences of any 
accident.
    The proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in the 
evaluation of radiological consequences.
    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.
    Fuel oil is not an accident initiator. Therefore, the 
possibility of a new or different kind of accident will not be 
created in relationship to the proposed changes to the TS. No 
modifications are proposed to the existing fuel oil storage system 
that would alter the design function or the ability of the DG to 
perform its safety function.
    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 to the 7-day time dependent fuel oil volume 
results in an increase in volume to accommodate fuel oil consumption 
needed to support the EPU.
    The reduced volume associated with the 5-day full load volume is 
equivalent to less than one hour of runtime and does not result in a 
significant reduction in a margin of safety because the 
calculational method results in a conservative estimate of the 
amount of fuel that would be needed during a design bases accident.
    The proposed change does not result in a change of the design 
bases for the DG or its support systems. The system will continue to 
provide a reliable source of power for safe shutdown of the reactor, 
assuming the single failure of one of the DGs. Independence, 
redundancy, and testability are maintained such that the required 
safety function can be performed by either DG train.

[[Page 70717]]

    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: N.S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Michael K. Webb, Acting.

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

    Date of amendment request: November 5, 2004.
    Description of amendment request: Waterford 3 Technical 
Specification (TS) currently requires all the Dry Cooling Tower (DCT) 
fans with cooling coils under the missile grating to be operable during 
a tornado watch. If one (or more) of these DCT fans is inoperable 
during a tornado watch, it is required to be restored to operable 
status within one hour or place the plant in Hot Standby within 6 
hours. The purpose of this TS change is to allow the plant to take 
credit for the DCT fans that are not under the missile grating to meet 
the fan requirements specified in TS Table 3.7-3. In addition, the 
proposed change will delete the requirement to monitor ambient 
temperature conditions when the DCT fan is inoperable on an inoperable 
train of the Ultimate Heat Sink (UHS).
    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 will delete the requirement to have all the 
DCT fans with cooling coils under the missile grating operable 
during a tornado watch. It has been determined (using tornado 
missile strike probability methodology--TORMIS) that the probability 
of damage to the DCT components not under the missile grating (fans, 
motors, associated conduits, electrical boxes, and cooling coils) is 
acceptably low. With respect to the probability of occurrence or the 
consequences of an accident previously analyzed in the FSAR [Final 
Safety Analysis Report], the possibility of a tornado reaching 
Waterford 3 and causing damage to plant systems, structures and 
components, including the DCT fans, is a design basis event 
considered in the FSAR. The probability of a tornado-generated 
missile strike on the DCT components was analyzed using the NRC 
[Nuclear Regulatory Commission] Staff approved probability method 
TORMIS. TORMIS showed that the change from essentially relying on 
DCT fans with cooling coils under the missile grating to relying on 
all operable DCT fans during a tornado watch is acceptable and 
represents an acceptably low probability of occurrence of tornado 
generated missile strikes on the DCTs. On this basis, the proposed 
change is not considered to constitute a significant increase in the 
probability of occurrence or the consequences of an accident.
    The proposed change to TS Action 3.7.4.d eliminates an 
unnecessary requirement, to determine ambient conditions and verify 
compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan 
is inoperable due to its associated train of UHS being inoperable. 
The determination of ambient temperature conditions and validation 
of the required number of fans based on the temperature will 
continue to be required when an UHS fan is inoperable and the 
associated train of UHS is operable. The UHS fans will not dissipate 
the required heat load when the associated train of UHS is 
inoperable, assuming the coincident ambient wet bulb temperature (78 
[deg]F) at the historically highest ambient dry bulb temperature 
(102 [deg]F). This change represents a burden reduction and has no 
impact on plant safety. This change also does not impact the 
initiators or mitigation of any design basis event.
    The proposed revision to TS Table 3.7-3 ensures consistency with 
the revisions to the TS Actions. This change is administrative and 
has no impact on the initiators or the mitigation of accidents 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will delete the requirement to have all the 
DCT fans with cooling coils under the missile grating operable 
during a tornado watch. It has been determined that the probability 
of damage to the DCT components not under the missile grating is 
acceptably low. A tornado at Waterford 3 is a design basis event 
considered in the FSAR. Therefore, the change will not contribute to 
the possibility of or be the initiator for any new or different kind 
of accident, or occur coincident with any of the design basis 
accidents in the FSAR. The low probability threshold established for 
tornado missile damage to system components is consistent with these 
assumptions.
    The proposed change to TS Action 3.7.4.d eliminates an 
unnecessary requirement, to determine ambient conditions and verify 
compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan 
is inoperable due to its associated train of UHS being inoperable. 
The determination of ambient temperature conditions will continue to 
be required when an UHS fan is inoperable with the associated train 
of UHS operable. There are no plant modifications or design changes 
proposed.
    The proposed revision to TS Table 3.7-3 ensures consistency with 
the revisions to the TS Actions. This is an administrative change.
    The above changes also do not have any impact on plant systems 
nor do they have any impact on the way plant systems are operated. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident.
    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. The existing licensing basis for Waterford 3 with 
respect to the design basis event of a tornado reaching the plant, 
generating missiles, and directing them toward the DCT components is 
to provide positive missile barriers. The basis for the proposed 
change recognizes that there is a low probability, below an 
established acceptance limit, that a tornado missile will strike DCT 
components. The change from essentially relying on DCT fans with 
cooling coils under the missile grating to relying on all operable 
DCT fans during a tornado watch is acceptable and represents an 
acceptably low probability of occurrence of tornado generated 
missile strikes on the DCTs. Therefore, this change is not 
considered to constitute a significant decrease in the margin of 
safety.
    The proposed change to TS Action 3.7.4.d eliminates an 
unnecessary requirement, to determine ambient conditions and verify 
compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan 
is inoperable due to its associated train of UHS being inoperable. 
The determination of ambient temperature conditions will continue to 
be required when an UHS fan is inoperable with the associated train 
of UHS operable. When the UHS is not available, the fans cannot 
dissipate the required heat load, assuming the coincident ambient 
wet bulb temperature (78 [deg]F) at the historically highest ambient 
dry bulb temperature (102 [deg]F). Therefore, it is not necessary to 
monitor ambient temperature and ensure the fan requirements of TS 
Table 3.7-3 are met when the UHS train is inoperable. This change 
represents an operational burden reduction and has no impact on 
plant safety.
    The proposed revision to TS Table 3.7-3 ensures consistency with 
the revisions to the TS Actions. These changes are administrative 
and have no impact on the operation of the plant, mitigation of 
analyzed events, or plant safety,
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, Entergy concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.


[[Page 70718]]


    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: N.S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Michael K. Webb, Acting.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket 
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 
2, Rock Island County, Illinois

    Date of application for amendment request: February 27, 2004, as 
supplemented by letter dated October 11, 2004.
    Description of amendment request: The proposed amendments would 
revise the Dresden Nuclear Power Station and Quad Cities Nuclear Power 
Station technical specifications (TS) to add the Oscillation Power 
Range Monitor (OPRM) instrumentation to 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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    This proposed change has no impact on any of the existing 
neutron monitoring functions. It activates the OPRM scram function 
and updates the TS to add the OPRM-related functions.
    Activation of the OPRM scram function will replace the current 
methods that require operators to insert an immediate manual reactor 
scram in the reactor operating region where thermal hydraulic 
instabilities could potentially occur. While this region will 
continue to be avoided during normal operation, certain transients, 
such as a reduction in reactor recirculation flow, could place the 
reactor in this region. Operation in this region, with the OPRM 
instrumentation scram function activated would no longer require an 
immediate manual scram and thus may potentially cause a marginal 
increase in the probability of occurrence of an instability event. 
This potential increase in probability is acceptable because the 
OPRM function will automatically detect the instability condition 
and initiate a reactor scram before the Minimum Critical Power Ratio 
(MCPR) Safety Limit is reached. Consequences of the potential 
instability event are reduced because of the more reliable automatic 
detection and suppression of an instability event, and the 
elimination of dependence on the manual operator actions. Operators 
will continue to monitor for indications of thermal hydraulic 
instability when the reactor is operating in regions of potential 
instability as a backup to the OPRM instrumentation.
    The potential for spurious reactor scrams has been evaluated. 
Operating experience with the OPRM has not resulted in the 
generation of any spurious reactor scram signals.
    Therefore, the proposed changes do not involve a significant 
increase in the probability of 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 changes replace procedural actions that were 
established to avoid operating conditions where reactor 
instabilities might occur with an NRC approved automatic detect and 
suppress function (i.e., OPRM).
    Potential failures in the OPRM trip function could result in 
either failure to take the required mitigating action or an 
unintended reactor scram. These are the same potential effects of 
failure of the operator to take the appropriate action under the 
current procedural actions. The effects of failure of the OPRM 
equipment are limited to reduced or failed mitigation, but such 
failure cannot cause an instability event or other type of accident.
    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?
    The OPRM trip function is being implemented to automate the 
detection and subsequent suppression of an instability event prior 
to exceeding the MCPR Safety Limit. The OPRM trip provides a trip 
output of the same type as currently used for the APRM [Average 
Power Range Monitor]. Its failure modes and types are identical to 
those for the present APRM output. Since the MCPR Safety Limit will 
not be exceeded as a result of an instability event following 
implementation of the OPRM trip function, it is concluded that the 
proposed change does not reduce the 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: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of amendment request: October 4, 2004.
    Description of amendment request: The proposed change requests 
approval to apply the Westinghouse best-estimate loss-of-coolant 
accident (BELOCA) analysis methodology to Beaver Valley Power Station, 
Unit Nos. 1 and 2, and requests an amendment of the related Technical 
Specifications. The BELOCA methodology has previously been approved on 
a generic basis by the NRC as presented in Topical Report WCAP-12945-P-
A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), ``Code 
Qualification Document for Best-Estimate LOCA Analysis,'' March 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?
    No. No physical changes are required as a result of implementing 
best-estimate large break loss of coolant accident (LOCA) 
methodology and associated Technical Specification changes. The 
plant conditions used in the analysis are bounded by the design 
conditions for all equipment in the plant. Therefore, there will be 
no increase in the probability of a LOCA. The consequences of a LOCA 
are not being increased, since it is shown that the emergency core 
cooling system is designed so that its calculated cooling 
performance conforms to the criteria contained in 10 CFR 50.46, 
Paragraph b. No other accident is potentially affected by this 
change.
    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 analyzed?
    No. There are no physical changes being made to the Beaver 
Valley Power Station units. No new modes of plant operation are 
being introduced. The parameters used in the analysis are within the 
design limits of the existing plant equipment. All plant systems 
will perform as designed during the response to a potential 
accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?

[[Page 70719]]

    No. It has been shown that the methodology used in the analysis 
would more realistically describe the expected behavior of plant 
systems during a postulated LOCA. Uncertainties have been accounted 
for as required by 10 CFR 50.46. A sufficient number of LOCAs with 
different break sizes, different locations and other variations in 
properties are analyzed to provide assurance that the most severe 
postulated LOCAs are addressed. It has been shown by analysis that 
there is a high probability that all criteria contained in 10 CFR 
50.46, Paragraph b are met.
    Therefore the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: October 15, 2004.
    Description of amendment request: The licensee proposed to revise 
Section 1.7, regarding the definition of ``Instrument Channel 
Calibration,'' of the Technical Specifications by incorporating the 
additional guidance for instrument channels containing resistance 
temperature detector (RTD) and thermocouple sensors provided by the 
``Standard Technical Specifications, General Electric Plants, BWR 
[Boiling-Water Reactor]/4 Specifications,'' NUREG-1433, Revision 3. The 
revised definition would permit in place qualitative assessment of the 
RTDs and thermocouples, and to allow a signal to be injected downstream 
of the sensor for the purpose of calibrating the remainder of the 
channel.
    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 revises the definition of Instrument Channel 
Calibration to allow RTD and thermocouple sensors to be 
qualitatively assessed with the remainder of the channel being 
calibrated normally. Instrument channel calibration is not an 
initiator of any accident previously evaluated. Furthermore, the 
proposed change will not affect the ability of the channel being 
calibrated to respond as assumed in any accident previously 
evaluated. The qualitative evaluation of sensor behavior for non-
adjustable sensors will provide an accurate indication of sensor 
operation and will assure that portion of the channel is operating 
properly, ensuring that the consequences of an accident will remain 
as previously evaluated. Therefore, the proposed Technical 
Specification 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 change revises the definition of Instrument Channel 
Calibration to allow RTD and thermocouple sensors to be 
qualitatively assessed with the remainder of the channel being 
calibrated as at present. 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. The proposed change also does not adversely 
affect the operation or operability of existing plant equipment. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change involves the definition of Instrument 
Channel Calibration to allow RTD and thermocouple sensors to be 
qualitatively assessed with the remainder of the channel being 
calibrated normally. The proposed change to the Instrument Channel 
Calibration definition does not alter the ability of a channel to 
respond as designed or as assumed in the safety analyses. 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 22, 2004.
    Description of amendment request: The licensee proposed to revise 
the Technical Specifications (TSs), Section 5.0, ``Design Features,'' 
by relocating the information to the Updated Final Safety Analysis 
Report (UFSAR). Specifically, the amendment would relocate these 
Sections: 5.3, ``Reactor Vessel,'' 5.4, ``Containment,'' and 5.6, 
``Seismic Design.'' The licensee stated that such information does not 
meet the criteria of 10 CFR 50.36(c)(4) for inclusion in the TSs. The 
information to be relocated to the UFSAR already exists in the UFSAR, 
and will continue to be controlled by 10 CFR 50.59 and 10 CFR 50.71(e).
    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 relocates certain design details from the TS 
to the UFSAR, where the information already exists. The UFSAR is 
maintained in accordance with 10 CFR 50.71(e). Any future change to 
these design details as described in the UFSAR will be evaluated per 
the requirements of 10 CFR 50.59 to assure that the change does not 
result in more than a minimal increase in the probability or 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    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. The 
proposed change will not impose or eliminate any requirements, and 
adequate control of the information will be maintained in accordance 
with applicable regulatory requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change has no impact on any analysis assumptions. 
The design details that

[[Page 70720]]

are being removed from the TS already exist in the UFSAR. Any future 
change to these design details described in the UFSAR will be 
evaluated per the requirements of 10 CFR 50.59 to assure that the 
change does not result in a design basis limit [or] a fission 
product barrier being exceeded or altered.
    Therefore, the proposed change does not involve a significant 
reduction in [a] margin of safety.

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

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

    Date of amendment request: October 14, 2004.
    Description of amendment request: The proposed changes correct 
administrative errors in Technical Specifications 3.10.i and 6.9.a.4.A.
    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?
    NMC [Nuclear Management Company, the licensee] Response for 
Proposed Change to TS 3.10.i: No. The NMC has reviewed the proposed 
change in accordance with the provisions of 10 CFR 50.92 to show no 
significant hazards exist. This change is being proposed to correct 
an administrative error that currently exists within the KNPP 
[Kewaunee Nuclear Power Plant] Technical Specifications; therefore 
it would not have an affect on the probability of an accident 
previously evaluated.
    NMC Response for Proposed Change to TS 6.9.a.4.A: No. The NMC 
has reviewed the proposed change in accordance with the provisions 
of 10 CFR 50.92 to show no significant hazards exist. This change is 
being proposed to correct an administrative error that currently 
exists within the KNPP Technical Specifications; therefore it would 
not have an affect on the probability 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?
    NMC Response for Proposed Change to TS 3.10.i: No. The proposed 
change does not alter plant configuration, operating setpoints, or 
overall plant performance. Therefore, the proposed change would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    NMC Response for Proposed Change to TS 6.9.a.4.A: No. The 
proposed change does not alter plant configuration, operating 
setpoints, or overall plant performance. Therefore, the proposed 
change would 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?
    NMC Response for Proposed Change to TS 3.10.i: No. The proposed 
change does not involve a significant reduction in a margin of 
safety. Inclusion of the omitted word ``and'' in TS 3.10.i will 
enhance the margin of safety.
    NMC Response for Proposed Change to 6.9.a.4.A: No. The proposed 
change does not involve a significant reduction in a margin of 
safety. Correction of the references in TS Section 6.9.a.4.A will 
enhance 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: July 9, 2004.
    Description of amendment request: The Humboldt Bay Power Plant 
(HBPP), Unit 3, is a decommissioning nuclear power plant that was 
permanently shutdown in July 1976. In December of 2003, Pacific Gas and 
Electric (PG&E or the licensee) applied for a license to store its 
spent fuel in an onsite dry cask independent spent fuel storage 
installation (ISFSI). Moving the spent fuel to an ISFSI would permit 
the licensee to begin significant decommissioning activities. The 
licensee has chosen to use a Holtec HI-STAR HB spent fuel cask handling 
system involving a spent fuel multipurpose canister and overpack. To 
facilitate spent fuel transfer from the HBPP spent fuel pool to the 
ISFSI, the licensee will also need to install a new crane that can be 
used to lift the cask handling system loaded with spent fuel 
assemblies. The licensee states it will be able to satisfy the 
applicable guidance of NUREG-0612, ``Control of Heavy Loads at Nuclear 
Power Plants,'' and NUREG-0554. ``Single-Failure Proof Cranes for 
Nuclear Power Plants,'' in performing the necessary movement of the 
HBPP spent fuel to dry cask storage. The licensee has requested a 
license amendment that approves the use of the crane and associated 
changes to the HBPP Defueled Safety Analysis Report (DSAR) along with 
analyses, design, and procedural changes required to implement transfer 
of the spent fuel from the spent fuel pool to the ISFSI.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. With the HI-STAR HB System and the associated design and 
handling procedures, all cask drops and other events, which could 
damage other spent fuel, have been precluded through the robust 
handling systems, and mechanical arrangement that preclude crane 
movement over spent fuel, meeting the guidelines of NUREG-0612. 
Revisions of the HBPP procedures implementing the control of heavy 
loads ensures that PG&E will meet the NUREG-0612 guidelines and will 
protect the fuel storage locations and the new HI-STAR HB System 
loading/unloading activities. As a result of this design approach, a 
cask-handling accident that results in a significant offsite 
radiological release is not considered credible as demonstrated by 
the probabilistic evaluation that was performed using the guidelines 
of NUREG-0612 Appendix B and updated information from NUREG-1774 
[``A Survey of Crane Operating Experience at U.S. Nuclear Power 
Plants from 1968 through 2002.'']
    Other HBPP licensing-basis events, such as the drop of a spent 
fuel assembly, have not been affected by these changes and remain 
bounding events for potential radiological consequences.
    The proposed design of the dry cask system, the handling system, 
and associated procedural controls provide assurance that: (1) 
Operational errors and mishandling events, and (2) support system 
malfunctions will not result in an increase in the probability or 
consequence of an accident previously analyzed.
    The proposed changes to use the Holtec HI-STAR HB system have 
been evaluated for seismic events and tornado missile impacts and it 
has been determined that these changes will not result in an 
increase in the probability or consequences of an accident 
previously evaluated. The Fire Protection Program will ensure that 
the combustible materials are properly controlled such that the 
total combustibles meet the current program commitments. Therefore, 
the

[[Page 70721]]

proposed changes do not involve a significant increase in the 
probability or consequences of an accident.
    2. Does the proposed amendment create the possibility of a new 
or different type of accident from any accident previously 
evaluated?
    No. The engineering design measures and the handling procedures 
preclude the possibility of new or different kinds of accidents. 
Damage to 10 CFR 50 structures, systems, and components from the 
cask handling and associated activities, and events resulting from 
possible damage to contained fuel have been considered. Both the 
types of accidents and the results remain within the envelope of 
existing HBPP DSAR licensing basis analyses, as demonstrated by the 
PG&E and Holtec analyses.
    The rupture of multipurpose canister (MPC) dewatering, forced 
helium dehydration or related closure system lines or the 
malfunction of equipment during cask handling operations resulting 
in radiological consequences are bounded by the HBPP DSAR fuel-
handling accident analysis.
    Other design considerations, such as spent fuel pool (SFP) 
thermal, water chemistry and clarity, criticality, and structural, 
were evaluated and determined not to introduce the possibility of a 
new or different kind of accident from any previously evaluated.
    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 amendment involve a significant reduction 
in a margin of safety?
    No. With the Holtec HI-STAR HB System, and the associated design 
and handling procedures, cask drops and other events have been 
precluded through robust load handling systems, providing defense-
in-depth as described in NUREG-0612. Cask tipovers, while not 
considered credible, are shown to be below the 60g limit, preventing 
damage to the contained fuel assemblies (and associated structures), 
and meeting the analysis guidelines of NUREG-0612. As the existing 
licensing basis assumes a nonmechanistic drop damaging the SFP and 
all fuel, the result of this design approach with the minimization 
of drops and the associated structural challenges assure the margin 
of safety has been maintained.
    Other HBPP licensing-basis events, such as the drop of a spent 
fuel assembly, have not been affected by these changes and remain 
bounding events. Revision of HBPP procedures implementing the 
control of heavy loads to incorporate the additional restrictions on 
heavy loads movement will not affect the procedures or methodology 
used and will, therefore, not affect margins.
    Adverse effects from seismic events and/or cask drops or 
tipovers have been evaluated, assuring that the fuel, MPC, and 
overpack remain within their design bases. Since design basis 
criteria are fully satisfied, there is no impact on the margin of 
safety.
    The Fire Protection Program will continue to ensure that the 
combustible materials are properly controlled such that the total 
combustibles meet the current program commitments. Thus, there are 
no significant reductions in margin of safety associated with these 
changes.
    Other design considerations, such as SFP thermal, water 
chemistry, criticality, and structural, were evaluated and 
determined to not involve a reduction in a margin of safety.

    Based on the above evaluations, the licensee concludes that the 
activities associated with the above changes present no significant 
hazards consideration under the standards set forth in 10 CFR 50.92 and 
accordingly, a finding by the NRC of no significant hazards 
consideration is justified.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esquire, Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Claudia Craig.

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

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments 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 in 
license amendment applications in the Federal Register on April 15, 
2003 (68 FR 18294). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 8, 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 one valve inoperable instead of requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the 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.

    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, Luzerne County, Pennsylvania

    Date of amendment request: September 22, 2004.
    Description of amendment request: The proposed amendment would 
extend

[[Page 70722]]

the validity of the reactor pressure vessel (RPV) pressure-temperature 
(P-T) limit curves from May 1, 2005, to May 1, 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The evaluation for the Unit 2 P-T limit curves for 32 EFPYs 
[effective full-power years] was performed using the approved 
methodologies of 10 CFR [Part] 50 Appendix G and Code Case-640. The 
curves generated from these methods were approved as Amendment 174 
(Ref. 1) and are currently in the Unit 2 TS. These curves ensure the 
P-T limits will not be exceeded during any phase of reactor 
operation. Resolution of the current industry issues related to 
fluence calculation methodology required PPL to limit applicability 
of the curves to May 1, 2005 for Unit 2. The proposed change does 
not alter any of the technical information shown on the present P-T 
curves. The change extends the expiration date for one year while 
maintaining the total accumulated exposure well below the 32 EFPY 
maximum exposure lifetime limit. Therefore, there is no increase in 
the probability or consequences of any previously evaluated accident 
as a result of this change.
    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 involves changing the expiration date on the 
Unit 2 P-T limit curves. The change does not affect the present 
operating margin in the P-T limit curves for inservice leakage and 
hydrostatic pressure testing, non-nuclear heatup and cooldown, and 
criticality. Operation in accordance with the present P-T curves, 
developed in accordance with the provisions of ASME Code [American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code], 
Section XI, Appendix G; 10 CFR [Part] 50 Appendix G, and ASME Code 
Case-640 provides adequate protection against a non-ductile-type 
fracture of the RPV. This proposed change does not create the 
possibility of any new or different [kind] of accident. The change 
extends the expiration date of the present P-T curves and does not 
result in any new or unanalyzed operation of any system or piece of 
equipment important to safety.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The technical information contained in the present P-T curves 
approved by Amendment 174 (Ref. 1) is not affected by this change. 
Extending the expiration date of the curves from May 1, 2005 to May 
1, 2006 will not reduce the margin of safety to RPV brittle 
fracture.
    Since the Unit 2 P-T curves have a maximum lifetime exposure of 
32 EFPYs and the anticipated exposure by May 1, 2006 will be well 
below the maximum value, the margin of safety is not reduced as the 
result of this change in expiration date. Resolution of the current 
industry issues related to fluence calculation methodology requires 
PPL to limit applicability of the Unit 2 P-T curves to May 1, 2006.

    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.

Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: September 23, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to require automatic starting 
of the auxiliary feedwater (AFW) pumps upon trip of the Turbine Driven 
Main Feedwater (TDMFW) pumps only when one or more of TDMFW pumps are 
operating.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The design basis events which impose AFW safety function 
requirements are loss of normal main feedwater, main feedline or 
main steamline break, loss of offsite power, loss of coolant 
accident, and small break loss of coolant accident. These accident 
evaluations assume actuation of AFW occurring due to low-low steam 
generator level or a safety injection signal. These signals are 
required safety related features unlike start-up of the AFW pumps 
due to the trip of both TDMFW pumps which is an anticipatory 
function and not required for either transient or accident analyses. 
Requiring this function only when the TDMFW pumps are running will 
not impact any previously evaluated design basis events. 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?
    No. This TS change involves the automatic start of the AFW pumps 
when the TDMFW Pumps trip. This change involves a function that is 
not a safety related feature and, therefore, is not credited in 
either transient or accident analyses. Since this change only 
affects the point at which this trip function needs to be operable 
and does not affect the function that actuates AFW due to low-low 
steam generator level or a safety injection signal, it will not be 
an initiator to a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in 
margin of safety?
    No. This TS change involves the automatic start of the AFW pumps 
when the TDMFW pumps trip which is not a safety related plant 
function. This change does not change any values or limits involved 
in a safety related 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: October 27, 2004.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.7.3, ``Main Feedwater Isolation Valves 
(MFIVs),'' to add the main feedwater regulating valves (MFRVs) and the 
associated MFRV bypass valves (MFRVBVs). In addition, the allowed 
outage time, or completion time, for inoperable MFIVs would be 
extended.
    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 add the MFRVs and MFRVBVs to TS 3.7.3 and 
extend the

[[Page 70723]]

Completion Time for one or more MFIVs inoperable from 4 hours to 72 
hours. Extending the Completion Time is not an accident initiator 
and thus does not change the probability that an accident will 
occur. However, it could potentially affect the consequences of an 
accident if an accident occurred during the extended unavailability 
of the inoperable MFIV. The increase in time that the MFIV is 
unavailable is small and the probability of an event occurring 
during this time period which would require isolation of the MFW 
[main feedwater] flow paths is low. Moreover, the redundancy 
provided by the MFRVs and MFRVBVs, which have the same actuation 
signals and closure time requirements as the MFIVs, provides 
adequate assurance that automatic feedwater isolation will occur if 
called upon.
    Therefore, the proposed changes do 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 accident previously 
evaluated.
    Closure of the MFIVs is required to mitigate the consequences of 
the Main Steam Line Break and Main Feedwater Line Break accidents. 
The MFRVs and MFRVBVs provide a diverse backup to this function. 
[The extended Completion Time for inoperable MFIVs is not an 
accident initiator.] The proposed changes do 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 changes do not revise any Technical Specification 
[Safety] Limit or accident analysis assumption. Therefore, [they do] 
not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: October 27, 2004.
    Description of amendment request: The amendment would delete or 
revise license conditions in the operating license for the Callaway 
Plant because the requirements are either obsolete or adequately 
described elsewhere. The amendment would also revise Technical 
Specification Tables 5.5.9-2, ``Steam Generator Tube Inspection,'' and 
5.5.9-3, ``Steam Generator Repaired Tube Inspection,'' to delete the 
requirement to notify the NRC pursuant to 10 CFR 50.72(b)(2) if the 
steam generator tube inspection results in a C-3 classification because 
reporting requirements are given in the regulations.
    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.
    This request involves administrative changes only. The changes 
consist of duplicates or overly burdensome reporting requirements or 
the deletion of completed items required by [the TSs or] conditions 
from the original issuance of Operating License NPF-30 [for the 
Callaway Plant]. No actual plant equipment or accident analyses will 
be affected by the proposed changes. Therefore, the proposed changes 
do 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 accident previously 
evaluated.
    This request involves administrative changes only. The changes 
consist of duplicates or overly burdensome reporting requirements or 
the deletion of completed items required by [the TSs or] conditions 
from the original issuance of Operating License NPF-30. No actual 
plant equipment or accident analyses will be affected by the 
proposed change[s] and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, the proposed changes 
do not create a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, Reactor 
Coolant System pressure boundary, and containment structure 
[pressure boundary]) to limit the level of radiation dose to the 
public. This request involves administrative changes only.
    No actual plant equipment or accident analyses will be affected 
by the proposed change[s]. The changes consist of duplicates or 
overly burdensome reporting requirements or the deletion of 
completed items required by [the TSs or] conditions from the 
original issuance of Operating License NPF-30. Additionally, the 
proposed changes will not relax any criteria used to establish 
safety limits, will not relax any safety system settings, or will 
not relax the bases for any limiting conditions of operation. 
Therefore, the proposed changes do 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: September 15, 2004.
    Description of amendment request: The proposed changes will change 
the Administrative Controls Section of the Technical Specifications 
(TS) in order to incorporate title changes, change the location where 
the plant-specific titles and TS titles are correlated, and relocate 
the unit staff requirements to the Quality Assurance Program. These 
proposed changes will support the implementation of proposed Virginia 
Electric and Power Company Topical Report DOM-QA-1, ``Nuclear Facility 
Quality Assurance Program Description,'' currently under NRC staff 
review. In addition, these proposed TS changes eliminate the 
descriptions of the onsite and offsite safety review organizations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of Surry Units 1 and 2 in accordance with the 
proposed license amendments would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change is administrative in nature and does not 
affect plant systems, structures or components (SSCs) or plant 
operation during normal or accident conditions. The proposed change 
only affects the designated titles of personnel, rewords or 
relocates requirements within TS or deletes requirements that are 
either not required to be part of TS or are already required by 
regulation. The change also relocates the detailed description of 
the onsite and offsite safety review organizations and non-licensed 
personnel qualification requirements to the Quality Assurance 
Program. Therefore, this change has no bearing on the probability of 
an accident. The management organizational structure and safety and 
operational reviews have not changed and, therefore, do not impact 
the ability of operating procedures or administrative controls to 
prevent or mitigate

[[Page 70724]]

a previously evaluated accident. As such, this change does not alter 
the conclusions of the existing safety analyses and therefore does 
not alter the consequences of an accident previously evaluated.
    2. Operation in accordance with the proposed license amendments 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed administrative change continues to ensure that 
adequate management oversight exists at the plant in accordance with 
the existing Technical Specifications. The proposed change only 
affects the designated titles of personnel, rewords or relocates 
requirements within TS or deletes requirements that are either not 
required to be part of TS or are already required per regulation. 
The change also relocates the detailed description of the onsite and 
offsite safety review organizations and non-licensed personnel 
qualification requirements to the Quality Assurance Program. 
Therefore this change does not impact plant SSCs or plant operation 
and therefore does not create the possibility of an accident of a 
different type than evaluated previously. The management 
organizational structure and safety and operational reviews have not 
changed. Therefore, there is no change in the method of plant 
operation, operation review or system design review. There are no 
new or different accident scenarios, accident initiators, nor 
failure mechanisms that will be introduced due to this change.
    3. Operation in accordance with the proposed license amendments 
would not involve a significant reduction in a margin of safety.
    The proposed change only affects the designated titles of 
personnel, rewords or relocates requirements within TS or deletes 
requirements that are either not required to be part of TS or are 
already required per regulation. The change also relocates the 
detailed description of the onsite and offsite safety review 
organizations and non-licensed personnel qualification requirements 
to the Quality Assurance Program. Consequently, this change does not 
impact plant design, plant operation or any safety margin and, 
therefore, does not significantly reduce a margin of safety.
    This evaluation concludes that the proposed amendments to the 
Surry Units 1 and 2 Technical Specifications do not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, do not create the possibility of a 
new or different kind of accident and 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: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: Mary Jane Ross-Lee (Acting).

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

    Date of amendment request: July 23, 2004.
    Description of amendment request: The amendment would revise 
Technical Specification 3.6.3, ``Containment Isolation Valves,'' by (1) 
Adding the abbreviation ``(CIV)'' for containment isolation valve in 
Condition A of the Actions for the Limiting Condition for Operation; 
(2) deleting the Note and revising Condition A to be for only one 
penetration flow path with one CIV inoperable; (3) revising the 
completion time for Required Condition A.1 from 4 hours to as much as 7 
days depending on the category of the inoperable CIV; and (4) revising 
Condition C to be for two or more penetration flow paths with one CIV 
inoperable. The proposed amendment is based on Topical Report WCAP-
15791-P, ``Risk-Informed Evaluation of Extensions to Containment 
Isolation Valve Completion Times.''
    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 to the Completion Times do not 
change the response of the plant to any accidents and have an 
insignificant impact on the reliability of the containment isolation 
valves. The containment isolation valves will remain highly reliable 
and the proposed changes will not result in a significant increase 
in the risk of plant operation. This is demonstrated by showing that 
the impact on plant safety as measured by the large early release 
frequency (LERF) and incremental conditional large early release 
probabilities (ICLERP) is acceptable. These changes are consistent 
with the acceptance criteria in [the risk-informed] Regulatory 
Guides 1.174 and 1.177. Therefore, since the containment isolation 
valves will continue to perform their [safety] functions with high 
reliability as originally assumed and the increase in risk as 
measured by LERF and ICLERP is acceptable, there will not be a 
significant increase in the consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended [safety] function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Further, the proposed changes do not increase the types 
or amounts of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed changes are consistent with the 
safety analysis assumptions and resultant consequences [in Chapter 
15, ``Accident Analysis,'' of the Updated Final Safety Analysis 
Report (USAR) for the plant].
    Therefore, it is concluded that this change does not increase 
the probability of occurrence of a malfunction of equipment 
important to safety.
    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 do not result in a change in the manner in 
which the containment isolation valves provide plant protection. 
There are no design changes associated with the proposed changes. 
The changes to Completion Times do not change any existing accident 
scenarios, nor create any new or different accident scenarios.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
consistent with the safety analysis assumptions and current plant 
operating practice.
    Therefore, the possibility of a new or different malfunction of 
safety related equipment is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do 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 impacted by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis. The calculated impact on risk is insignificant and is 
consistent with the acceptance criteria contained in Regulatory 
Guides 1.174 and 1.177.
    Therefore, it is concluded that this 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

[[Page 70725]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert Gramm.

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

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

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendments: October 29, 2004.
    Brief description of amendments: Provide a one-time change to 
Function 4a, ``Reactor Coolant System (RCS) Hot Leg Temperature 
Indication,'' of Technical Specification Table 3.3.4-1. This would 
allow continued operation until the next refueling outage (spring of 
2005) with one out of four RCS hot leg temperature indications 
inoperable in the Auxiliary Control Room.
    Date of publication of individual notice in the Federal Register: 
November 5, 2004 (69 FR 64596).
    Expiration date of individual notice: November 19, 2004.

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, 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 NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by e-mail to [email protected]. (Note: Public 
access to ADAMS has been temporarily suspended so that security reviews 
of publicly available documents may be performed and potentially 
sensitive information removed. Please check the NRC Web site for 
updates on the resumption of ADAMS access.)

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

    Date of application for amendment: August 27, 2004, as supplemented 
by letters dated October 11 and 19, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specifications, Section 2.1.A, changing the safety limit minimum 
critical power ratio value from 1.09 to 1.10 for both four-or five-
recirculation-loop operation, and from 1.10 to 1.12 for three-
recirculation-loop operation.
    Date of Issuance: November 16, 2004.
    Effective date: November 16, 2004, and shall be implemented within 
60 days of issuance.
    Amendment No.: 252.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 14, 2004 (69 
FR 55467). The October 11 and 19, 2004, letters provided clarifying 
information within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of 
this amendment is contained in a Safety Evaluation dated November 16, 
2004.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, et. al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station (OCNGS), Ocean County, New Jersey, Docket 
No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin 
County, Pennsylvania

    Date of application for amendments: March 23, 2004, as supplemented 
June 16, 2004.
    Brief description of amendments: The amendments relocate the 
Independent Onsite Safety Review Group requirements from the 
Administrative Controls in Section 6 of the Technical Specifications to 
the Exelon Generation Company, LLC (EGC)/AmerGen Energy Company, LLC 
(AmerGen) Quality Assurance Topical Report (QATR) at TMI-1 and OCNGS. 
In addition, administrative corrections are included, which update 
references to the EGC/AmerGen QATR, which has replaced the OCNGS and 
TMI-1 Operational Quality Assurance Plans.
    Date of issuance: November 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 251 and 252.
    Facility Operating License Nos. DPR-16 and DPR-50: Amendments 
revised the Technical Specifications.
    Date of initial notices in Federal Register: May 11, 2004 (69 FR 
26186).
    The supplement dated June 16, 2004, 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 determinations.
    The Commission's related evaluation of the amendments is contained 
in a

[[Page 70726]]

Safety Evaluation dated November 8, 2004.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: April 23, 2004.
    Brief description of amendment: The amendment deletes Technical 
Specification Section 6.16, ``Post-Accident Sampling Programs NUREG 
0737 (II.B.3, II-F.1.2),'' and the related requirements to maintain a 
Post Accident Sampling System.
    Date of issuance: November 22, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment No.: 253.
     Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
26187)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 22, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 12, 2004, and 
supplemented by letters dated June 16 and September 2, 2004.
    Brief description of amendments: The amendments modify the LaSalle 
Technical Specifications (TS) to eliminate selected response time 
testing requirements associated with Reactor Protection System 
instrumentation and Primary Containment Isolation instrumentation for 
Main Steam Line Isolation functions. Specifically, the changes revise 
the response time testing requirements for TS Section 3.3.1.1, 
``Reactor Protection System (RPS) Instrumentation,'' Reactor Vessel 
Steam Dome Pressure--High function and TS Section 3.3.6.1, ``Primary 
Containment Isolation Instrumentation,'' Reactor Vessel Water Level--
Low Low Low, Level 1 and Main Steam Line Pressure--Low functions.
    Date of issuance: November 19, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 169, 155.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the TS.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19569).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 23, 2003.
    Brief description of amendment: The amendment modifies technical 
specification (TS) requirements to adopt the provisions of Industry/TS 
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode 
Restraints.''
    Date of issuance: November 10, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 219.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 16, 2004 (69 
FR 55844).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 10, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 14, 2004.
    Brief description of amendment: The amendment relocates the 
requirements of Technical Specification 3.3(1)a, ``Reactor Coolant 
System and Other Components Subject to ASME XI Boiler & Pressure Vessel 
Code Inspection and Testing Surveillance'' and TS 3.4, ``Reactor 
Coolant System Integrity Testing,'' to the Updated Safety Analysis 
Report (USAR). Requirements in TS 3.3(1)a were related to inservice 
inspection of ASME Class 1, 2, and 3 components and requirements in TS 
3.4 were related to reactor coolant system integrity testing.
    Date of issuance: November 8, 2004.
    Effective date: November 8, 2004, and shall be implemented within 
120 days from the date of issuance.
    Amendment No.: 230.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 22, 2004 (69 FR 
34703)
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated November 8, 2004.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: June 22, 2004, as supplemented 
on September 27, 2004.
    Brief description of amendments: The amendments revise the 
frequency associated with Surveillance Requirement (SR) 3.3.8.1.4, 
which directs the performance of the logic system functional test, from 
once every 18 months to once every 24 months. The amendments change the 
SRs in Hatch, Units 1 and 2 Technical Specifications.
    Date of issuance: November 22, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 243/186.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Surveillance Requirements in the Technical 
Specifications.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46592).
    The supplement dated September 27, 2004, provided clarifying 
information that did not change the scope of the June 22, 2004, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 22, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: October 29, 2004, as supplemented 
November 5, 2004.
    Description of amendment request: The amendment provides a one-time 
change to Function 4a, ``Reactor Coolant System (RCS) Hot Leg 
Temperature Indication,'' of Technical Specification (TS) Table 3.3.4-1 
to allow continued operations until the next refueling outage with one 
out of four RCS Hot Leg Temperature Indications inoperable in the 
Auxiliary Control Room.

[[Page 70727]]

    Date of Issuance: November 19, 2004.
    Effective date: As of the date of issuance and shall be implemented 
immediately upon receipt.
    Amendment No.: 53.
    Facility Operating License No. (NPF-90): Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. On November 5, 2004, the Commission issued a 
notice (69 FR 64596) that included the staff's proposed determination 
that the amendment request involves no significant hazards 
consideration (NSHC). The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received. The notice also provided an opportunity to request 
a hearing by November 19, 2004, but indicated that if the Commission 
makes a final NSHC determination, any such hearing would take place 
after issuance of the amendment. The supplement of November 5, 2004, is 
within the scope of that notice, and did not change the proposed no 
significant hazards consideration.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated November 19, 2004.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

    Dated at Rockville, Maryland, this 29th day of November, 2004.
    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-26606 Filed 12-6-04; 8:45 am]
BILLING CODE 7590-01-P